IAEA-TECDOC-699
Impact of extended burnup on the nuclear fuel cycle
Proceedings of an Advisory Group Meeting held Vienna,in December2-5 1991
INTERNATIONAL ATOMIC ENERGY AGENCY
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Printed by the IAEA in Austria April 1993 FOREWORD
Currently, burnup extension is one of the most advanced and intensively studied areas in fuel technology and management worldwide. The average design discharge burnups that commercialle ar y availabl range BWRr eth fo PWR d n ei san 35-4e sar 5- MW-d/k40 d an gU MW-d/k0 5 , respectivelygU . Economic incentive exisy r extendinma sfo t g burnup even further t leasa MW-d/k0 o t ,6 t . gU
Burnup extension affects several important stages of the fuel cycle and concerns the whole nuclear fuel cycle industry. Increase of burnup reflects on the requirements for natural uranium, enrichment and fuel fabrication, reactor core configuration and its control, fuel performance and the back end of the fuel cycle, such as spent fuel handling, transportation, treatmen storaged an t .
In view of the importance of high burnup for water reactor fuel utilization, the IAEA convened technical committee meetings (in 1981, 1984 and 1990) in the framework of the International Working Grou Waten po r Reactor Fuel Performanc Technologd ean y which consists of representatives from 24 countries and three international organizations. Additionally, around 50% of the presentations at the IAEA Symposium on Improvements in Water Reactor Fuel Technology and Utilization in 1986 were related to subjects on high burnup.
The Water Reactor Fuel Extended Burnup Study (WREBUS) started in 1988 and was completed in 1991. It was the first internationally conducted study of its kind involving eleven IAEA Member States. The findings of the study have been published as IAEA Technical Reports Series No. 343 in 1992.
These proceedings present national approache extendeo st d burnu experiencd pan e subjecte outlind th n an i evaluatiot n ea ,ar futur e state th th f ef eo n o trend fielde th .n si
The IAEA wishes to thank all the participants of the Advisory Group Meeting for their valuable contribution and especially the Chairman, Mr. P. Lang, and the Session Chairmen, Messrs . DuflouP . . . GriffithJ Hesketh,K e d IAEan Th sA . officer responsible th r fo e organizatio meetine th f preparationd o gan f thino s documen. SukhanovG . Mr s i t , Nuclear Materials and Fuel Cycle Technology Section, Division of Nuclear Fuel Cycle and Waste Management. EDITORIAL NOTE
In preparing this material for the press, staff of the International Atomic Energy Agency have mounted and paginated the original manuscripts as submitted by the authors and given some attention to the presentation. The views expressed in the papers, the statements made and the general style adopted are the responsibility of the named authors. The views do not necessarily reflect those of the governments of the Member States or organizations under whose auspices the manuscripts were produced. thisin The bookuse of particular designations countriesof territoriesor does implynot any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention specificof companies theirof or products brandor names does implynot any endorsement recommendationor IAEA. partthe the of on Authors are themselves responsible for obtaining the necessary permission to reproduce copyright material from other sources. Throughout textthe names Memberof States retainedare theyas were when textthe was compiled. CONTENTS
SUMMAR ADVISORF YO Y GROUP MEETING
1. INTRODUCTION ...... 7 .
. IMPAC2 EXTENDEF TO FUEE DTH L F BURNUFRONCYCLE 7 O D TH . ETN . EN P O
2.1. Impac extendef o t d burnu resourcn po e utilization ...... 7 . 2.2. Impac highef o t r enrichmen fronfacilitien d o t en t s ...... 8 . 2.2.1. Light water reactors ...... 8 2.2.2. Heavy water reactors ...... 8 . 2.3. Impact of fuel design changes required for extended burnup on front end facilities ...... 8 2.4. Impac f extendeo t d burnuqualite th f n plutoniupyo o recycled man d uranium an associaten do d facilities ...... 8 .
3. IN-REACTOR ISSUES ...... 9 .
3.1. Nuclear design ...... 9 . 3.1.1. Impact on core characteristics ...... 9 3.1.2. Impact on nuclear fuel design ...... 9 3.1.3. Impac fuen o t l management strategie loadind san g patterns ...... 0 1 . 3.1.4. Operational flexibility ...... 10 3.1.5. Extensio codef no extendeo st d burnup conditions ...... 0 1 . 3.2. Fuel performance ...... 0 1 . 3.2.1. Corrosion ...... 10 3.2.2. Fission gas accumulation and release ...... 11 3.2.3. Pellet clad interaction ...... 11 3.2.4. Fuel reliability ...... 1 1 . 3.3. Mechanical design ...... 2 1 . 3.3.1. Structural design impacts ...... 12 3.3.2. Impact on materials ...... 12 3.4. Thermalhydraulics ...... 12 3.5. Mixed oxide fuels ...... 12
. IMPAC4 EXTENDEF TO FUEE TH LDF 3 CYCL BURNUO 1 BAC E D . TH KE.. EN N PO
4.1. General issues ...... 13 4.1.1. Criticality ...... 3 1 . 4.1.2. Heat output ...... 13 4.1.3. Radioactivity ...... 13 4.1.4. Accurac f calculateyo d actinid fissiod ean n product parameters ...... 3 1 . 4.2. Storage ...... 3 1 . 4.3. Transport ...... 14 4.4. Reprocessing ...... 4 1 . 4.5. Spent fuel conditioning and disposal in once-through cycle ...... 14
5. CONCLUSIONS ...... 14 PAPERS PRESENTE ADVISORE TH T DA Y GROUP MEETING
Development slightln si y enriched uraniu power mfo r reactor fue Argentinn i l a ...... 9 1 . J.A. Casario, LA. Alvarez Extended burnu CANDpn i U ...... 3 2 . P.O. Boczar, J. Griffiths, I.J. Hastings Necessity for and feasibility of burnup extension in China's nuclear industry ...... 33 Zhenglun Liu Present status of the nuclear fuel cycle in the CSFR ...... 37 F. Pazdera Near-term plans towards increased discharge burnups in Finnish power reactors: Review of background aspects and current activities ...... 44 KelppeS. Compatibility of extended burnup with French fuel cycle installations ...... 49 P. Demoulins, P. Saverot Extensio burnuf no Indiapn i n PHWRs ...... 7 5 . H.D. Purandare, P.D. Krishnani, K.R. Srinivasan Burnup extension plan of BWR fuel and its impact on the fuel cycle in Japan ...... 64 A. Toba High burnups for water reactor fuels - A UK perspective ...... 83 K. Hesketh Effect f extendeso d burnunucleae th n po r fuel cycle ...... 7 8 . P.M. Lang Improvemen fueR fuel e WE effecburnus f th it l o tcycln d o t pan e ...... 1 9 . Yu.K. Bibilashvili, G.L. Lunin, V.D. Onufriev, V.N. Proselkov Burnup as a safety parameter for handling transport packages and storage facilities for spent nuclear fuel ...... 99 V.l. Kulikov, N.S. Tikhonov, G.Ya. Philippov, A.I. Ivanyuk nucleae th f o Impac bacre d fueth ken n lo t cycle ...... 4 10 . H. Bairiot
List of Participants ...... 111 SUMMAR ADVISORF YO Y GROUP MEETING
1. INTRODUCTION
The Advisory Group Meeting was held in Vienna from 2 to 5 December 1991, to review, analyse discusd an , effecte sth f burnuso p extensio botn i h lighheavd an t y water reactor l aspectal fue e n so th l f cycleso . Twenty experts from thirteen countries participated in this meeting. consensuThera s ewa s that both economi environmentad can l benefite sar driving forces toward the achievement of higher burnups and that the present trend of bumup extensio expectee b y nma continueo dt economie Th . c aspect r LWRsfo s have been assesse resulte th d sd an publishe IAEe th Adn i Technical Report 1992n i s3 Serie34 . . sNo
Batch average bumup levels were considered to be feasible goals, achievable within the next 10-20 years. For light water reactors burnup would be extended to 55-60 GW-d/t U for PWRs and WWER 1000 reactors and to 45-50 GW-d/t U for BWRs and WWER 440 reactors. For heavy water reactors burnup could be extended to 12-20 GW-d/t U. The extended burnup has been considered for the three main stages of the fuel cycle: the front end, in-reactor issues and the back end.
2. IMPAC EXTENDEF TO FUEE DTH LBURNU FRON F E CYCLO TH D N TPEEN O
The following potential effects of extended bumup on the front end of the fuel cycle have been identifie discussedd dan :
impac resourcn o t e utilization, impact of higher enrichments on front end facilities and operations, impac f necessaro t y fuel design modification fronn facilitiesd so en t , case th f close eo n i d cycles, impac f extendeo t d burnuqualite th n f plutoniupyo o m uraniud an m recovered after reprocessin thein o rd subsequengan t utilization.
2.1. Impact of extended burnup on resource utilization
e extentioTh f burnuo n p leada significan o t s t decreas f specifio e c uranium consumption (expressed in grams of uranium used per MW-h of generated power) and to a reductio f fueno l assembly specific consumption (numbe f assemblieo r fabricatee b o st r dfo givea n energy produced) LWRsr Fo .magnitud e ,th decrease th f eo uraniuf eo m consumption burnue th y pdependwa extensioe th n so carries ni d out mors i t i : e effectiv case th f eo n ei unchanged cycle length than when burnup and cycle lengths are simultaneously increased.
The effect on consumption of separate work units [SWUs] (number of SWUs used per unit of energy generated) is less clear (the change can be either or slight increase or decrease according to the burnup change) but estimates are that it would be small.
The issue of impact on resource utilization has been treated on a quantitative basis for LWR Technican si l Reports Serie 343. sNo .
No complementary work has been found necessary on this subject in the short term.
7 2.2. Impac f higheo t r enrichmen fron facilitien d o t en t s
2.2.1. Light water reactors
The use of higher LJ-235 enrichment for use in LWRs has possible impacts upon safety for enrichment plants, fuel fabrication plants, transport containers and fresh fuel storage facilities.
y Thirequir ma sexpenditure eth f timefford eo ean documentation i t d minonan r modification in order to meet the licensing requirements, but essentially there were no generic problems envisaged. existinMose th f to g enrichmen fued ltan fabrication facilitie alreade sar y licensed for enrichments up to 5% U-235 which is consistent with the framework of extended burnup under consideration. This conclusion remains valid bearin minn gi d tha defininn i t g the enrichment limit required, the necessary safety margin to allow for uncertainties needs to consideree b thad dan t whe averagn na e enrichment leve stateds i l , some fuel management strategie requiry sma e enrichments abov belod ean w that average.
There may be generic issues related to adapting front end facilities to enrichments much greater than 5% U-235, however within the framework of the range of extended burnup considered by this group, such issues were not considered relevant.
2.2.2. Heavy water reactors
One option for achieving burnup extension for HWRs is the use of slightly enriched uranium plac(SEDn i f natura)e o l uranium sucn I . caseha currene th , t license limitr sfo natural uranium fuel fabricators would reviewedneee b do t additionn I . , typical storagd ean transport designs and procedures for fresh fuel have to be reviewed to verify criticality aspects.
An alternative way of obtaining SEU is to utilize reprocessed uranium from LWR fuels. This could imply modifications of fabrication facilities and special QC and dose control procedures in order to cope with the associated activity.
2.3. Impac fuef o t l design changes require extender dfo d burnup on fron facilitied en t s
Extended burnup may require changes to fuel rod and assembly designs, and to fuel management strategies (see impacSectioe Th . t n3) upo n fronfacilitied en t s arises from: f integrao e us l e burnablTh e) absorber(a reactivitr sfo y control; (b) Change materialn si desigd san fuer nfo l rodassembliesd san .
Such new requirements could imply additions and modifications in current manufacturing processe proceduresC Q d san . However participante th , t identifno y d sydi an major problems associated with achieving these requirements somr Fo . e plant changeo sn s needee other fo ar d r plantdan s changes have already been made.
2.4. Impac f extendeo t d burnu qualite th n f plutoniupyo o m and recycled uranium and on associated facilities
Extended burnup can change significantly the isotopic composition of the plutonium uraniud an m recovered after reprocessing, enhancing their radioactivit increasind yan e gth content in neutron absorbing isotopes.
8 s feli It t that such degradatio qualite th f f recycleno o y d uranium could require additional protective measure fronfacilitiee d th en t n si s wher materiae eth treates i l d before being reuse reactoa dn i r (enrichment facilities, manufacturing chainf o se devoteus e th do t REU, transport containers). However no need for significant additional modification of such facilitie anticipates i envisagee th r dfo d burnup extension frame situatioe Th . n could become more critica casn i l f multipleo e recyclin same th f ego material; however thi clearls i a t yno short or medium-term issue.
In the case of recycling plutonium in MOX, the effect of the degradation of quality may give rise to significant dose uptake issues with possible implications on the plant design, unless remote operation and eventually maintenance are realized. This needs further consideration.
Attention should als paie ob possiblo dt e impacpresence th f o t f suceo h active materials on decommissioning methods or cost of the relevant facilities.
Another issu tha es i highee tth r leve absorbinf o l g isotope recyclee th n si d materials leads to consequent even higher enrichments for LWRs, which in turn result in additional licensing problems.
. 3 IN-REACTOR ISSUES
3.1. Nuclear design
3.1.1. Impact on core characteristics
In all reactor types, extended burnups are achieved by higher initial fissile isotope concentrations increasee Th . d thermal neutron absorptio associatee th d nan d hardeninf go the neutron spectrum combine to influence the core physics characteristics in a number of ways. At the fundamental level main considerations are: firstly, fission products tend to decrease reactivity with increasing burnup offse,e whicb usiny y b t hma g burnable absorbers. e potentiaOn l impac o reducf t thao ts i t e shutdown margins. Secondly e reactivitth , y coefficients are modified. Thirdly, especially in the absence of burnable absorbers, the reactivity difference between new fuel and previously burned fuel is increased with extended burnups, potentially causing increased radial peaking factors.
The impact of extended burnup operation on the nuclear design characteristics are complicate many db y factors detaileA . d discussio f thesno e factor outsids si scope eth f eo this summary importann a t bu , t poin noto t t thas ei t ther tendenca s ei modifo yt fuee yth l management scheme loadind san g pattern same th t esa tim goins a e higgo t h burnupd san this complicate discussione sth importans i t I . beao t minn i r d this point when examinine gth impac f higo t h burnups impace .Th loadinn o t g pattern discusses i d furthe Section i r n 3.1.3.
3.1.2. Impact on nuclear fuel design
The principal impact on the nuclear design of extended burnups and associated fuel management changes is the need to use higher enrichments. In LWRs this is associated with larger reactivity investment starfuee a f th lo t cyclsa e which lead requirementura n si o nt t for more reactivity hold-down from control rods, soluble boron or burnable absorber. In PWRs WWERsd an highee th , r beginnin f cyclgo e hold-down requirement neeresuly e th sma dn i t burnable tous e absorber ensuro st e thabeginnine th t f cyclgo e moderator temperature coefficient remains within the design limit. BWRs may also require increased burnable absorber loadings, due to the limited reactivity hold-down that the control rods can achieve. With respect to burnable absorbers the higher loadings needed with extended burnups gives an increased incentive to use integral fuel absorber designs. There are many integral fuel absorber designs availabl t presentea PWRn I . s boron coating gadolinid san usee aar d extensively at present. In BWRs gadolinia is used exclusively. Both these types and future variant suitable wite sar us h r extendeefo d discharg f integraeo burnupse us l e fueTh .l burnable absorbers impacts on the fuel management and fuel performance and is discussed furthe Section i r n 3.3.
There is a general tendency at present to increase the number of fuel rods in PWR assemblie bundlesR R BW HW d trend a , an san d which exists independentl trene th f dyo towards extended burnups. This will increase thermal design margins thao ,s t within-assembly peaking factor constraints may be relaxed. Thus the trend towards extended burnups will not necessarily result in any additional increase in the complexity of nuclear fuel designs.
3.1.3. Impact on fuel management strategies and loading patterns
The trend towards extended burnups tends to be accompanied by an increase in complexity of the fuel management strategies. Along with the trend towards extended burnups many LWR utilities are inclined to adopt low leakage loading patterns. These have several advantages, suc reduces ha d vessel fluenc improved ean d fuel utilization t generallbu , y require more sophisticated core loading patterns. Thus radiae th , l peaking factors tene b do t higher in low leakage cores, necessitating the use of burnable absorbers. The use of low leakage loading patterns in conjunction with extended burnup operation can thus lead to higher burnable absorber requirements. With respect to vessel fluence, extended burnup operatio advantageoue b n nca s becaus opportunite eth highese yth arisee us t burnuso t p fuel assemblies to shield the reactor vessel.
3.1.4. Operational flexibility
An important consideration of adopting extended discharge burnup operation is that the operational flexibility availabl plane th teo t operator affectede b y sma . Thu LWRn si e sth higher radial peaking factors associated with extended discharge burnup cycles, especiallyn i conjunction witleakagw hlo e loading patterns reducy ma margin,e eth s availabl thermao et l design limits and, depending on the specific details of a particular plant, may give rise to reduced operating flexibility.
Fuel management strategie keeo t s wilpm ai withil n existing constraints thao e s , th t impac f extendino t g discharge burnup operationan so l flexibilit minimizeds yi .
3.1.5. Extension codesof extendedto burnup conditions
Computer codes develope validated dan corr dfo e calculations extendeneee b o dt d qualified an extender dfo d burnup conditions ensuro ,t e that calculations done with such codes do not represent extrapolations but are normalized against data and therefore can be considered fully reliabl futurr efo e extended burnup applications.
3.2. Fuel performance
3.2.1. Corrosion
e issu Th f clao e d (external) corrosio hign i n h discharge burnup operatios nwa identifie beins da mose gth t important fuel performance issu PWRsr efo , particularl plantyn i s with high coolant temperatures. Although clad corrosio potentiaa ns i l concer other nfo r reactor significana t e seeb no o nt typess t wa issut i , t presentea . Thu differene sth t clad material
10 use WWEdn i R reactors showb N d s, an exceptionallbase allor n Z a f yn o d o y good corrosion performance. Similarly BWRn i , s clad corrosiomajoa t no r issues ni choice Th .f claeo d material in relation to PWR clad corrosion is discussed in Section 3.3.2.
3.2.2. Fission gas accumulation and release
With extended burnups ther greatea s ei r accumulatio f fissiono n gase mord san e opportunity for diffusion processes to release the gas from the fuel pellets. However, this does not necessarily resul highen i t r partial pressures from fissio releass nga thers trenea a es i d BWRn i HWRd san s towards increasin numbee gth fuef ro l rod assemblr spe thereford yan e reducing the linear generation rate independent of the trend to extended discharge burnups. However, wit hfixea d numbe rodf assemblyr o spe certainls i t i , case yth e that higher fission releases ga obtainee sar d with extended burnup mak y thid ss ma ean necessart i introducyo t e rod design changes, as discussed in Section 3.3.1. The increased fission gas release will give additional incentive to improve the fission gas release predictive capabilities of fuel performance codes.
An important point regarding the impact of extended burnups is that although the inventory of long lived fission products is increased, the inventory of short lived fission products is essentially unaffected, as they are at equilibrium. The impact of high discharge burnups on the short term radiological consequences of accidents, being dominated by the short lived volatile fission products therefors ,i e minimal.
At extended burnup advantagese b ther y ema movinn si g away fro presene mth R tLW design requirement that fuel rod internal pressures must not exceed the coolant pressure. It might prove feasible to replace this deterministic design criterion with a probabilistic argument t higa hd burnupan s there woul more db e incentiv implemeno et t suc changeha .
A more complete account of the effect of extended discharge burnups on fission gas release can be found in the reports of the IAEA meeting on fuel performance held at Studsvik and ANS/ENS LWR Fuel Performance Conference held at Avignon, April 1991.
3.2.3. Pellet clad interaction
Whil trene eth d towards extended burnup mossn i tsignifican o casen s sha t impacn o t the susceptibility of fuel clad to damage from pellet clad interaction, there is nevertheless an indirect impact resulting fro fuee mth l management changes associated with high burnups. Los f operationaso l flexibilit connection yi n with load follow operatio PWRn i identifies swa d aren a f particulaas o a r concer somn i e countries.
With regard to BWRs and HWRs, however, the trend towards higher ratings is offset by the trend towards increasing the number of fuel rods per assembly or bundle making this consideratio f lesno s importance.
Possible materials changes to improve PCI performance are discussed in Section 3.3.2.
3.2.4. Fuel reliability
Available evidence point thero t s e bein adverso n g e impac f extendeo t d burnup operatio fuen no l reliability. Indee l fued al alread s i l y designe licensed dan operatior dfo n well below the levels at which life limiting phenomena occur. Moreover there is some evidence of there bein possiblga e benefit understooe .b Thin sca d fro reducee mth d rat t ewhica h fresh feed assemblie loadee sar extenden di d burnup operation, giving reduced ris f fueko l defects related to manufacturing defects.
11 3.3. Mechanical design
3.3.1. Structural design impacts
As mentioned previously, ther trena es i increas do t numbee eth fuef ro l rod PWRn si , assembliesR HW BW d Ran . This decreas fuee th l n elineai r powe temperaturd an r e allows an increase in burnups. Other design changes can be also used e.g. to adopt a hollow pellet which decreases the fuel center temperature and therefore the FGR and the plenum volume.
3.3.2. Impact on materials
Clad external corrosion is a limiting factor for PWRs at the highest burnups and highest coolant temperatures. Ther therefors ei stronea g incentiv develoo et p corrosion resistant cladding materials. Possible solutions currently under consideratio r developmenno e ar t variants on the current Zircaloy alloys, duplex claddings in which a low corrosion alloy coats outee th r surfac Zircalof eo thu, y4 s obtainin corrosioga n resistant material while retainine gth bulk properties of Zircaloy 4.
With respect to pellet clad interaction and the underlying stress corrosion cracking mechanism, which is generally believed to promote crack growth, various materials properties developments are being pursued. The use of softer fuel pellets (such as niobia doped pellets), possibly in conjunction with hollow pellets is another approach being considered. The use of zirconium lined cladding is a well established means of enhancing pellet clad interaction margins in BWRs. Graphite coated clad is used in HWR bundles and zirconium liner clad is also being considere assemblieR HW r bundlesdd fo san .
3.4. Thermalhydraulics
In relation to extended discharge burnups, there are no direct impacts on the thermalhydraulics. Assembly design changes introduced in connection with extended discharge burnup operation or otherwise, must be evaluated against thermal hydraulic compatibility considerations. Design change thin si s category includ increasee eth d number of fuel rod r assemblspe r bundlyo e being considere r BWRd fo HWRd san possibld san e changes in fuel to moderator ratio being considered for some BWRs for extended discharge burnup operation highee Th . r power peaking factors associated with extended discharge burnup operation tend reduco st e margin thermao st providy l ma limit incentivn d ea san eo t implement design change improvo st e hydraulic mixing.
3.5 Mixed oxide fuels
fuelX thermasn i MO f o e l reactorus e Th s alter core sth e characteristic whicy wa ha sn i is qualitatively similar to increasing the uranium enrichment. Thus MOX fuel of equivalent lifetime average reactivity has higher thermal neutron absorption than UO fuel. This reduces
the reactivity worth of control devices, modifies the reactivity coefficients an2 d in LWRs requires Pu concentration zoning to control the within-assembly peaking in MOX assemblies at the interface assemblies2 withUO . These consideration t neee wele b no s ar do lo t knowd d an n elaborated here, since the use of MOX assemblies in LWRs is well established. The implementatio recyclX t establishedye MO t HWRn i f eno no s e presenci s X Th . MO f eo assemblie coreR LW s complicaten si fuee sth l management t givebu , n that modern loading patterns are commonly very heterogenous, this is not a major consideration.
In most respect issuee sth s relate extendeo dt d discharge burnup operatioe th e nar fuels2 UO .r Therfo s a evidencs e i X fuel X samlese MO sr ar eMO s e fo susceptiblI PC o et failures fuelstha2 thano UO consideratione s , th t s discussed unde havy r 3.2.ma loweea 3 r impact economice .Th recycl X improvee MO ef ar s o greatea do t wit2 r hthaUO extender nfo d
12 discharge burnups, because the dominant cost component is the fabrication cost, which is essentially independent of burnup. Another factor for MOX is that the isotopic composition of recovered Pu will change depending on the discharge burnups of the fuel from which the Pu is recovered. This has a minor impact on the nuclear design of MOX cores.
4. IMPACT OF EXTENDED BURNUP ON THE BACK END OF THE FUEL CYCLE
4.1. General issues
4.1.1. Criticality
e increasTh f initiaeo l fissile conten f fuer o extendet fo l d burnup will requirn ea assessmen f criticalito t y implications mosn I . t countrie reductio e credisa th r fo t reactivitn i y of fuel whic attaines hha designes dit d burnu allowedt no ps i . This positiorelaxee b y dnma witintroductioe hth f instrumentno s capabl f reliableo e measurement f eitheso r burnur po reactivity. In the case of fuel containing burnable absorbers the regulatory authorities in some countries allow credit to be taken for these absorbers.
4.1.2. Heat output
The heat output from extended burnup fuel with the same cooling time is higher due to the increased fission product and actinide inventories. Typical values for this increase are Frencgivee th n i h contributio proceedinge th no t thif so s meeting.
4.1.3. Radioactivity
The radiation fields due to the fission products increase almost in proportion to the burnup, whilactinidee th e o t thos e s edu increas e more rapidly. Typical e valueth r fo s increas radioactivitn ei founFrence e b th n dn i yhca contributio thino t s meeting.
4.1.4. Accuracy of calculated actinide and fission product parameters
Extended burnup fuel provides a potential source of an excellent data base which coul usee db validato dt e code prediction r compositionssfo , heat generatio radiatiod nan n fields. irradiatioe Prioth o rt thif no s fuel between-code comparison indicated sha d differences between the calculations of up to 30% for some nuclides.
4.2. Storage
Extended burnup results in a decrease in the rate of discharge of fuel assemblies. For a given storage capacity the time to use up that capacity is increased and the energy generated fro fuee mth l s greatei store t i r extenden dfo i r d burnup fuel. This provides somewhat longer cooling times without providing additional storage facilities.
While discharged extended burnup fuel is associated with higher radiation fields the water cover abov fue e existinr eth fo l storagt g we bee s eha n founadequate b o dt reduco et e these fields.
While wet storage provides adequate cooling for the increased heat generation of extended burnup fuel longer pre-cooling may be needed before dry storage can be used. Dry storage facilities may, if not already designed for extended burnup fuel, also require additional shielding.
13 4.3. Transport
Transport cask spenr sfo t fuelicensee ar l carro dt maximua yo t fue p u l m heat load dosd an e rate which depend upo burnue nth deca d pan y timeburnue Th . p limit differene sar t between countrie necessare b y rangd san GW-d/ma 0 t e5 I re-licens o y t o fro t . U 0 tm 3 r eo redesign transport casks or alternatively to use longer cooling times and/or partial cask loadings.
4.4. Reprocessing
The impact of extended burnup fuel on reprocessing can be viewed from the plant owners or the customers viewpoint. To the customer less volume of waste is returned for a given amoun f poweo t r generated specificatioe Th . waste th fina r f neo fo l disposal will stay withi guaranteee nth d targets approve safete th y ydb authorities e reprocessorth r Fo . , however, assuming constant throughput, extended burnup results in a slight increase in the waste from the reprocessing activities. This increase is due to the increased quantities of fission products, increased waste from solvent degradation and increased waste from sludge formation.
Existing reprocessing plants were designed for fuel up to a maximum burnup of 45 GW-d/t U and an enrichment of 4.5%. If, due to burn up extension, this maximum will be exceeded certain steps will need to be taken. Sufficient time would normally be available for these steps due to the approximately eight-year delay between the introduction of the extended burnup fuel intreactoeventuas e it o th d an r l arrivareprocessine th t a l g plant.
These steps include reassessmen: of t
criticality safety with particular emphasi thosn so e areas applying fissile concentration controls; effluent discharges; the heat removal capability; the capability of monitoring equipment to measure the increased activity of plutonium; the effects of reduced fuel solubility; operational dose uptake; the effect of new burnable absorber designs.
4.5. Spent fuel conditioning and disposal In once-through cycle
Prior to fuel disposal the fuel assemblies are placed in a containing system. The impacincreasee th n o t d heat generatio thin no s system potentiae th , l neutroe n th dos d ean shielding require increasee th r dfo d disposae fueth l n activito ld operatioan y n muse b t considered, if not avoidable by a number of other alternatives.
5. CONCLUSIONS
1. Exchange of information should be promoted on aspects such as licensing of extended burnup fuel, fuel design criteria and analysis of operational experience with extended burnup fuel.
2. Exchange of information on cross-section data for burnable absorbers is desired by some countries.
14 3. The programme to assess and benchmark computer codes for fuel thermal-mechanical performance at high burnups should be continued.
4. Measures to decrease uncertainties in predictions of isotopic composition and decay heat generatio f spenno t fuel shoul studiee db d further.
5. The IAEA is addressing aspects of concern primarily to the LWR community. Development are also occurring in the HWR community which need to be addressed. Calculational benchmark meetingscould be set up aimed at fuel behaviour and discharge fuel propertie highet sa r burnup reactivitd san y coefficients. Alternativele yth HWR equivalent of WREBUS could be prepared.
Next page(s) left blank 15 PAPERS PRESENTED AT THE ADVISORY GROUP MEETING DEVELOPMENT SLIGHTLSN I Y ENRICHED TABL ATUCHA-. E1 1 REACTOR TECHNICAL DATA______URANIU POWER MFO R REACTOR FUEL REACTOR TYP Epressurize: d heavy water reactor (PHWR) IN ARGENTINA THERMAL OUTPU : T117 W 9M . CASARIOJ.A . ALVARELA , Z GROSS ELECTRICAW L M OUTPU 7 36 : T Comision Naciona Energîe d l a Atômica, Buenos Aires, Argentina COOLANT AND MODERATOR : D2 O NUMBER OF FUEL CHANNELS : 253 Abstract NUMBE FUEF RO L ASSEMBLIE3 25 S: Argentina has two Pressurised Heavy Water Reactors (PHWK) in operation at the present time: one is Atucha-1 and the second Embalse. Both are REFUELLING AND FUEL SHUFFLING : ON-POWER natural uranium fuelled. The use of slightly enriched uranium (SEU) in PHWR is an effective way to increase the fuel discharge burnup. The main benefits from the use of SEU cores are the following: TABLE 2. ATUCHA-1 CORE TECHNICAL DATA_____ « Fuel cycle costs decrease • uranium resources savings FUEL TYP : Enatura l uranium dioxide « Spent fuel volume decreases. TOTAL NATURAL URANIUM INVENTORY : 39.0 t Other advantage corf so e enrichmen describee tar reporte th n di .
orden I teso overale rt tth l performance under irradiatio U fuelSE f ,a no QUANTIT FUER PE L FUE F D ASSEMBLYO LRO 6 3 : Y first serie twelvf so e Atucha- prototypeA 1F s (0.65% U-235s )wa QUANTITY OF SPACERS : fabricate Argentiny db e domestic supplier. ZRY-4 : 15 Work is carried out in Argentina in order to confirm the advantages and feasibility of PHWR fuel burnup extension programme. INCONE (botto1 : L m end)
. 1 INTRODUCTION TABLE 3. ATUCHA-1 FUEL ROD TECHNICAL DATA
The fuel discharge burnup of Atucha-1 Nuclear Power Plant can FUEL ROD OUTSID : E 11.DIAMETEm 9m R be increased by using slightly enriched uranium (SEU) in reload fuel . ACTIVE LENGTH OF FUEL ROD : 5300mm Hain characteristic e Atucha-th f o s 1 PWHe listear R n Tabli d . 1 e Ziy-: CLADDIN4 G MATERIAL The present Atucha-1 fuel data is summarized in Tables 2 and 3. AVERAGE FUEL ROD HEAT RATE : 232 W/cm e transitioTh n from natural uraniu U corm SE cor eo t ewoul e b d MAXIMU STATIONARR MFO Y CONDITIO W/c1 53 m : N obtained through two or three enrichment steps, in order to avoid changes to the thermalhydrauIic design of the core. MAXIMUM FOR NON-STATIONARY CONDITION : 600 W/cm The first step will be the use of 0.85Î U23S, followed by a second ste f abouo p t 1.00% e lasOn . t ste f abouo p t 1.ZO undes i Ï r MAXIMUM DESIG W/c0 N69 PEA m: K POWER consideration e see b n Tabl i n . ca A e s a , TABLE 4. ENRICHMENT PLANNED STEPS OF TABL . ATUCHA-E8 1 FUEL ASSEMBLY HOMOGENEOUS SEU ATUCHA-1 CORE U-236 Enrichment
. FIRST STEP : 0.85 % U-235 Natural 0.86 % 1.0 « 1.2 «
. SECOND STEP : 1.00 % U-235 AVERAGE DISCHARGE BURNUP tMWd/kgU] 6 11.4 16 21 AVERAGE RESIDENCE TIME [efpdl 198 377 529 694 . THIRD STEP : Not defined yet REFUELLING FREQUENCY [FA/efpdJ 1.28 o.67 0.48 0.36
3 11 8 14 8 20 6 39 QUANTIT J % F.A./yeaF Y5 O 8 • L [F r CONSUMPTIO9 3 URANIU0 F N4 O 4 4 M ItU/year 1 6 J TABL . MAIE5 N ADVANTAGE SSLIGHTLF FROO E MUS Y ENRICHED URANIUM ON ATUCHA-1 CORE
. EXTENSION OF FUEL DISCHARGE BURNUP . SAVINGS IN URANIUM RESERVES
. REDUCTION OF SPENT FUEL VOLUME TABLE 9. MAIN DESIGN GUIDELINES OF SEU ATUCHA-1 FUEL . LOWER TOTAL FUEL CYCLE COST . MAINTAIN THE FUEL ABILITY TO OPERATE RELIABLY TO EXTENDED BURNUPS
TABLE 6. OTHER CONSEQUENCES OF . AVOI INTRODUCTIOE DTH OPERATIOW NE F NO N CORE ENRICHMEN ATUCHA-1______N TO _ RESTRICTION POWEO ST R PLANT
. EXTENSION OF FUEL RESIDENCE TIME . KEEP THE PRESENT MARGINS OF SAFE OPERATION . REDUCTIO FUEF NO L ASSEMBLIES CONSUMPTIOD NAN OF THE REACTOR FRECUENC POWERN (O F YO ) REFUELLIN FUED GLAN SHUFFLING . REDUCTION OF FRESH FUEL STOCK AND FUEL TRANSPORTS . IMPACT ON SPENT FUEL STORAGE POOL CAPACITY
TABL . ADDITIONAE7 L ADVANTAGE CORF SO E TABLE 10. ENRICHMENT ON ATUCHA-1 DURING THE SEU ATUCHA-1 FUEL HIGHER PERFORMANCE TRANSITIO HOMOGENEOUO NT CORU SSE E REQUIREMENTS DUE TO . INCREASIN DISCHARGF GO E BURNUNATURAE TH F PO L AND LOWER ENRICHED URANIUM FUEL . HIGHER FUEL BURNUP
. POTENTIAL RELOADIN SPENF GO T NATURAL FUEL . LONGER RESIDENCE TIME ASSEMBLIES (STORED IN POOL) TO EXTEND THEIR BURNUPS . HIGHER FAST NEUTRON FLUENCE ATUCHA-U TABLSE . E11 1 FUEMAID NLRO LIFE LIMITING TABL . EMBALSE14 E FUEL ASSEMBLY______ASPECTS U-236 Enrichment
. RESISTANCE TO PCI STRESS CORROSION ASSISTED Natur«! 0.90 % 1.0 * 12 % DURING FUEL SHUFFLING AND CORE MANOEUVRINGS AVERAGE DISCHARGE BURNUP [MWd/kgU 5 7. 14.l 6 17.8 23.9 . INTERNAL PRESSURFUED LRO E AVERAGE RESIDENCE TIME [efpdl 303 590 720 966 . FUEL ROD DIAMETRAL CHANGES WITH RESPECT TO FUEL ROD-SPACER INTERACTION REFUELLING FREQUENCY [FA/efpdl 15.04 7.72 6.34 4.72
QUANTITY OF F.A./year IF • 85 %l 4665 2397 1966 1464 . ZIRCALOY WATERSIDE CORROSION L
TABLE 15. UTILIZATION OF DOMESTIC RESOURCES WITHI FUEU NSE L PROGRAM ARGENTINSN I A
ATUCHA-U TABLSE . E12 1 FUEL ASSEMBLY MAIN LIFE . FUEL DESIGN AND ENGINEERING CAPABILITY LIMITING ASPECTS . DEVELOPMENT OF NEW TECHNOLOGIES FOR MATERIALS . DIFFERENTIAL FUEL ROD AXIAL GROWING CONCERNING AND FUEL MANUFACTURING TO SPACER-BEARIN INTERACTIOD GPA N
. STRESS RELAXATION OF SPRING SLIDING SHOES RELATED TO COOLANT CHANNEL-FUEL ASSEMBLY CLAMPING FORCES
TABL . BENEFITE16 S FRO FUEU MSE L . FUEL CYCLE COSTS DECREASE
URANIUM RESOURCES SAVINGS
SPENT FUEL VOLUME DECREASE TABLE 13.
REMEDIES PLANNED TO PREVENT OR IMPROVE THE RESISTANCE TO PCI-SCC FAILURES OF The main advantages from the use of slightly enriched uranium on Atucha-1 core, are: SEU ATUCHA-1 FUEL - extensio f fueo n l discharge burnup, - savings in uranium reserves; . GRAPHITE COATING - reductio e spenth f to n fuel volum d lowean e r total fuel cycle cost. . PURE ZIRCONIUM LAYER These advantages are shown in Table 5. W Other consequences of core enrichment (see Table 6), are: U LIFSE E LIMITING ASPECTS N) - extension of fuel residence time, For Atucha-1 SEU fuel rods, the main life limiting aspects are - reduction of fuel assemblies consumption and lower frecuency (Table 11): f on-poweo r refuellin d fuean g l shuffling, -resistanc o pellet-claddint e g Interaction stress corrosion - reduction of fresh fuel stock and fuel transport; assisted (PCI-SCC) during fuel shuffling and core operational - Impact on spent fuel storage pool capacity. fluctuations, -internal fuel rod pressure, In addition, durin e transitioth g o homogeneut n D corSE s e there -fuel rod diametral changes with respect to fuel rod-spacer Is also an increase in discharge burnup and residence time of gap; the fuel with natural or lower enriched uranium. Reloading of spent natural fuel assemblie n ordei s o extent r d their original -waterside corrosion. w discharglo e burnu s alsI po under analysis. r fueFo l assemblies e principath , l life limiting aspecte ar s (Table 12): Table 8 compares the estimated average discharge burnups and other figures for different enrichments steps respect to our 235 - differential axial growin f fued o concerning ro l o t spacerg - current experience with natural uranium core r 0.85Fo . 1 U . sliding pad Interaction, - e fueth l consumptio n2 percen5 wil e b l t respece presenth o t t t value- - stress relaxatio f sprino n g sliding shoes related witd en h The annual savings of uranium are'above 28 percent considering a of life (EOl) clamping forces between coolant channel and tall assa r enrichmenfo y f 0,20o t % [)Z3!i an< ja plant load factor fuel assembly. 5 percent8 f o .
The reduction of refuelling frecuency represents a significant advantage for the extension of the fuelling machine lifetime and mantelnance operations. I REMEDIEPC . 5 S For higher enrichments and therefore higher burnups, Is of special concer e potentiath n f fueo l l failure y PCI-SCCb s . These failure e causeb y y poweb na ds r ramps during on-power FUEL DESIGN GUIDELINES fuel shuffling, core load change d coran s e startups o maintaiT . n satisfactory fuel performance without Introducing restrictions The Increases In average fuel assemblies discharge burnup and e presenth o t t fuel management d operatioan s n procedurese w , residence time wlJI require progressive changes of fuel rod and are planning to use PCI remedies like (Table 13): assembly designs. - graphite coatin f claddino g g Inner surface; e maiTh n design guideline U fueSE l r assembliefo s e (Tablar s: 9) e - pure Zirconium layes "barrier")(a r , bonde o claddint d g - Maintai e fueth nl abilit o operatt y e reliabl o extendet y d inner surface. burnups levels, - Avoi e introductioth d w powene f ro n operation restrictions; B yf thio usin e s on gremedies e expecw , o prevent t e th t - Maintain the present margins of safe operation of the reactor. decreasin e PCI-SCth f o gC failure threshol t highea d r burnups.
FUEL PERFORMANCE REQUIREMENTS 6. PROTOTYPES IRRADIATION Atucha-1 fuel assemblies with higher discharge burnup y usinb s g To start the Introduction of SEU fuel and for overall performan- SEU will have to meet increased performance requirements due e testingc a firs, t séri f twelvo e e Atucha-1 fuel assembly proto- to: types was fabricated by our domestic suppliers. These fuel assemblies contain oran.ium enriched to 0.85Î U235. - higher fuel burnup, - longer residence time; Design and specifications changes were introduced by our - higher fast neutron fluence (0t). engineerin a aesigan g ann dgroup desig, nini groups prototype. This prototypes wil e o irraaial s wilt l be irradiated e neath e nean r i th nf------r futur- «-..-L- - t Atucha-a e t n...-- - 1 »,_-^ Powe.____..,_.....^ r o approximatelPlant p u t . y s showI n Tabls i a n. 10 e 8.3 MWd/Kg.U. 7. SEU PROGRAMS FOR OTHER POWER REACTORS EXTENDED BURNU CANDPN I U operationn i P a CftHDU-NP Argentin s e o i seconTh . tw e 6s on d ha a PWHR at Embalse. P.G. BOCZAR . GRIFFITHSJ , , I.J. HASTINGS e fuer Th thifo l s reacto s alsi r o natural uranium. Chalk River Laboratories, Preliminary calculations were performed considering different AECL Research, levels of SEU for the Embalse reactor core. Chalk River, Ontario, The results are indicated in Table \b. Canada Reduction of fuel consumption and uranium resources savings are as important as i.t was stated for Atucha-t . Abstract We have also the intention to analyze the extension of SEU programs to the Atucha-2 Power Plant which is' under construction. The use of enriched fuel, from either enriched natural uranium or uranium recovered from LWR discharged fuel, is an attractive option CAHDe foth r U reacto enrichef o r e systemus de e impacTh th . f to uranium on resource utilisation and fuelling costs is discussed. . 8 DOMESTIC RESOURCES Some technical problems result and their solution ia described. We intend to develope these fuel SEU programs by using our domestic capability on fuel design (Table 15). The development of new manufacturing technology will be also carried out by our researchers and then transferred to our 1. BACKGROUND cladding and fuel manufacturing plants. CANDUe Th * reacto heavy-water-moderatea s ri -cooledd dan , pressure-tube reactor (Figureo t s1 3). The pressure tubes are arranged horizontally on a square lattice pitch (380 channels ia the CANDU 6 reactor). Each channel contains 12 natural-uranium UO-» fuel bundles, 0.5 m long. 9. CONCLUSIONS Fuelling is on-line, and bi-directional, meaning that adjacent channels are fuelled in the opposite direction. When a channel is refuelled, a fuelling machine attaches to each end of the channel: Analysi d calculationan s s performe o datt p eu dsugges t thae th t fresh fuel is inserted in one end of the channel, and old fuel that has reached its discharge burnup is benefits of discharge burnup extension in PHWR are: remove channele dth froCANDf a othe o e n md I th . ren Ureactor6 , abou channel2 t refuellee sar d usiny da 8-bundl n gr a pe e shift. This metho fuellinf do g give CANDe sth U reacto potentiae rth o lt - fuel cycle cost* decrease. accommodate a variety of different fuels. e preliminarOn y estimatio r Atucha-fo n 1 shows that wite th h first enrichment step (0.85$ s possibli t )i o savt e e approxima- tel0 millio1 y n dollar r year pe se fue Th . l contributioo t n On-line refuelling reduces the excess reactivity that needs to be controlled. Reactivity control generation cost decreases about 30 percent. (other than through refuelling) is primarily through adjuster rods, zone controller units, and shutoff rods. In a CANDU 6 reactor, there arc 18 adjuster rods, arranged in three rows. The adjuster - Uranium resources savings . rod verticalle sar y oriented betwee fuee nth l channels servd ,an e several purposes: For 0.85Î enriched Atucha-1 core, the Uranium savings are about 28 percent per year. • The adjusters provide xenon override time. After a reactor shutdown, the level of - Spent fuel volume decreases. xenon-135 increases for several hours. In a CANDU 6 reactor, the neutron absorption e amounTh f speno t t fuel reduceS r 0.85jfo & enriched Atucha-1, adjustere inth s balance absorptioe sth xenoe th n ni which build aften s i minute0 3 r s to 52 percent per year. afte shutdownra adjustee th f I . r rod withdrawe sar n afte trip a rreacto e e b ,th n rca restarted within 30 minutes—otherwise the reactor poisons out. These benefit e encouraginar s o continut s u g e with these programs » The adjusters provide shaping (flattening) of the flux and power distributions. r botfo h argentinean Nuclear Power Plants. • Withdrawa adjustesome f lo th f eo r rods provides reactivity even e shith mn ti thae th t fuelling machines are unavailable. • Withdrawa adjustesome f o lth f eo r rods provide reactivite sth y neede operato dt t lesea s than full power (power set-back step-backs)r so . ro CO CANDU: CANada Deuterium Uranium. Registered trademark. 3.8m mnel Row Designation ; ! ^ k. . l : î . ! ( : t ( « i' t: 1! v II t v 1l ^ > 2 Z z 1 . l l © © o o A © 0 s_/ D 0 ShutofO f Rods (28) i O o • • • 0 • • • • Adjuster Rods (21) © D 0 0 0 D O O Zone Control Units (6) © • o D Mechanical Control © a © 0 o D G Absorbers (4) r © © o /-> cy © o ^J o O e o m © O o o V
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24 savinge Th uraniun si CANDn i m U resultinSE U f dependo e enrichmenge frous th e n mso th f to bot produce hth t (th enrichmene tailee th fuel)th n s i d an , t plant saving(Figure Th . es4) t levea t ou l an enrichment of about 1.2%, and increase with decreasing tails enrichment. An enrichment of Fuel channels Insid notcf eo h wall 0.9% SEU results in annual uranium savings of between 20% and 30% relative to the use of natural uranium. With 1.2% SEU, the savings increase to between 30% and 40%. There is no resource incentiv goinn ei highego t r enrichments, whil technicae eth l difficulties increase with higher enrichments. (Note thaexace th t t savings depen somo dt e burnuextene th n to p assumer dfo natural uranium, which depends on several factors, such as the size of the reactor and the designed xenon override capability.) advantage Th e thanatural-uranium-fuellee tth d CANDU enjoys ove light-watee rth r reactor (LWR) oooooooooooooooooooooo oooooooooooooooooooooo uraniun i m utilizatio erodens i d with increasing butnup othed san r uranium-conserving fuel oooooooooooooooooooooo management techniques in the LWR. Moreover, a natural-uranium-fuelled CANDU could completely los traditionas eit l advantag uraniun ei m utilizatio AVLIw ne nf i S technology makes lower tails enrichments economical. WitfuelU h,SE CAND maintain Uca advantags nit n ei oooooooooooooooooooo uranium utilization alsn oca benefid ,an t from decrease enrichmenn si t prices tha expectee tar do t oooooooooooooooooo result from increased competition and new enrichment technologies (Figure 5). oooooooooooooooooo . oooooooooooooooo ooooooooooooo domestin i U SE cf o CAND e Whilus e Ueth reactors could reduce their annual uranium oooooooooooo requirement abouy sb t 30%effece th thif , o t s reductio Canadiae th n no n uranium mining industry Canada' f woullargee o b t 1988% n i d:no 85 s, outpu availabls twa e for export/5) Moreovere ,th CANDn i U domestiSE U f woulo e cus d creat edomestia c market that could suppor Canadiata n Inside of calandria wall enrichment industry/6) The export of enriched uranium, rather than natural uranium, would doubl value eth Canadiaf eo n uranium exports. Figure 3: Face View Of Reactor
zone-controe Th l system provides bulk spatial-reactivitd -an y control. This system consistf so vertical tubes containing either two or three compartments. Each compartment contains light water 50 (a neutron-absorber in CANDU), the level of which can be varied either uniformly in all the compartments (providing bulk-reactivity control), or individually (providing spatial control). Rapid shutdow CANDn i Uprovides i independeno tw y db t shutdown systems: shutoff rods, Tails whic normallc har y poised abov reactoe eth r liquid-poisocorea d ,an n injection systeme Th . 40 - reactivity depth and speed of insertion of both systems ensures effective shutdown capability in the While the natural-uranium-fuelled CANDU is the most neutron-efficient of all commercial reactors 20 in operation today, enriching the uranium improves the neutron efficiency even further/'-)4 the 1.0 1.2 1.4 1.6 1.8 increased natural uranium require provido dt source eth fissilf eo e U-23 enrichee th n 5i d fues i l more than compensate increasee th y b r dfo energy extracted fro fuelexampler e mth Fo . , while Enrichment (%) two kg of natural uranium are required to produce one kg of SEU with an enrichment of 1.2%, Figure 4: ro three time energe sth obtaines yi d fro fuele mth . Hence amoune th , naturaf to l uranium requireo dt Oi produce a unit of energy is reduced by a factor of 2/3, about 30%. Annual Uranium Saving sCandn I Witu huSe ro CD 250 2.1.2 Fuelling Costs >, D In existing CANDU reactors onle th , y area whicn si h significant cost reduction madn e i b e n ear sca § 0 PWR operation and maintenance, and in fuelling. SEU (and recovered uranium, RU, from spent LWR fuel) are the only fuel cycles in CANDU that are economic now, compared to natural uranium D 3.25% enrichment fuelling. Figure 6 shows the fuel-cycle cost savings (undiscounted) for SEU in CANDU, relative 200 -i a o 4.25%" natural-uraniuo t m fuelling representativr fo , e component costs. Ove widra e rang enrichmenf eo t and natural-uranium costs enrichmenn ,a f 1.2to near-optimal%s i onlt no ,y for uranium utilization, CANDU, ° but als fuel-cyclr ofo e costs. While result0.9U %SE annua n si l fuel-cycle cost saving betweef so n ————— B ————— Natural Uranium 20% and 25%, these savings increase to between 25% and 35% with an enrichment of 1.2%. The § 0 CANDU corresponding absolute savings amount to over $ 100 million a year in a system of 15 GWe 13 ISO- capacity/3) To A 0.9% enrichment r> c 7 9 A 4 1.2% " CANDU fuel-cycle costs are about a factor of two lower than for LWRs< - > - an important advantage, since this more than offset addee sth d heavy-water cost componen totae th ln uniti t » 1 energy cost. The use of SEU in CANDU will help maintain this advantage as LWR fuel-cycle cost reducee sar d through such improvement highes sa r burnu smalled pan r reload batches. 100 • 1 0. 2 0. 0.3 2.2 Recovered Uranium (RÜ) Tails Enrichmen) (% t 2.2.1 Uranium Utilization Figur: e5 The use in CANDU of uranium recovered from spent LWR fuel during reprocessing is attractive Comparison of CANDU and PWR from both the perspective of uranium conservation and fuelling costs. There is a strong resource Uranium Requirements incentive to recycle RU in CANDU, rather than re-enrich it for reuse in a pressurized-water reactor (PWR): twic thermae eth l energextractee b n yca whedU R fro ne mrecycleth CANDn di U rather 40 thaninanPWR.W«) 2.2.2 Fuelling Costs Costs, $/kg SWUU N , Fuelling of CANDU with RU would be economical today. A major benefit stems from not having to re-ennch the RU before using it in CANDU. This results in significant simplifications and cost- CO 30- D) o 100, 100 savings recenA . t stud shows yha economi e nth c benefi usinf o tdirectl U gR CANDyn i U rather lo • 1000 ,20 than re-enrichin PWR/a n i gt i > Figur CANDcompareed 7 an R Us PW fuelling costsr ,fo 'I representative cost assumptions. Fuelling costs are shown for two values of RU: RU at the same cos naturas ta l uraniu coso n t m a (RU2) (RU1) U fuelline R d Th . ,an g cost advantag CANDf eo U o 200, 100 extraordinarys i R ovePW e rth . CANDU fuelling costs with natural uraniu aboue mar factota f ro •3 20- • 200, 200 2 lower than PWR fuelling costs. With the cost of RU the same as natural uranium, PWR fuelling cost reducee sar Figuronln y i (PWRd b % 1 ywherea, 2 e7) RU , s CANDU fuelling cost nearlc sar y a factor of 3 lower than for the reference PWR (CANDU, RU1). With RU available at no cost, PWR fuelling costs decrease by 36% (PWR, RU2), whereas CANDU fuelling costs (CANDU. RU2lowefacto5 a 5. e f referencar )e ro th tha r nfo e PWR. 10 An upper value for RU can be derived by equating the cost of fabricating PWR fuel from RU to the 1.0 1.2 1.4 1.6 1.8 cost of fabricating PWR fuel from enriched uranium from natural-uranium feed: Enrichment (%) Cos enrichef to d uranium from natural-uranium feed (&1cg) = C CswC+ + I I F F u• • -SN Cuo+ U 2 Figure 6: U30UF68 Relative Fuel Cycle Cost Savings Cost of enriched uranium from RU (&1cg) CANDn i witU hUSE = CRU • F2 + (CUF6 + ACUF6) ' F2 + (Cswu + ACSWu) ' SRU + (CUO2 Using these value naturaa s, l uranium pric $50/kf eo resultgU /kgprica 0 n sRi f $4 ; ea o Uf o natural uranium pric $80/kf eo g $7&/kgf resulto pricU a R n si .f eo Hence PWRa n i , , recovered uranium has a value lower than that of natural uranium. There are several reasons for the cost advantage of using RU in CANDU over re-enriching it and using it in a PWR.<12> The RU from spent PWR fuel contains about 0.4% U-236, a neutron poison whose neutronic effect is an order of magnitude greater in the harder PWR spectrum than in CANDU. Re-enrichin increaseU R e onlt g th U-23 e sno y th 5 concentration alst U-23e ,obu th 6 concentration. Consequently, the U-235 must be enriched to a higher level to compensate for the presence of the U-236. (For example, a U-236 concentration in the RU of 1.2% necessitates an enrichment of about 3.6% U-235, compared to 3.25% U-235 using natural-uranium feed for the same burnup.) This higher enrichment increase requiremente U sth SW d an bot r U sfo h R (enrichment), and the associated costs. There are also cost penalties associated with the conversion, re-enrichment, and fabrication of PWR fuel using RU, due to the radioactivity of the daughter product U-232f so levee ,whicth f lo h also increases after re-enrichment. . 3 TECHNICAL CHALLENGES ASSOCIATED WITH EXTENDED BURNUP 3.1 Fuel-Management Considerations enrichmenf o e us e CANDn Th ti U would increas reactivite eth frese th hf y o fuel orden I . o rt reduc powee eth r ripple during refuelling, fewer bundles woul addede db exampler Fo . , with Figure 7: 0.9% SEU, which roughly double fuee sth l burnup fro mMWd/k7 g with natural uraniu4 1 mo t Comparison of CANDU and PWR Fuelling Costs MWd/kg 4-bundl,a e shift fuelling scheme coul usede db ; with 1.2% SEU, which triple fuee sth l bumup to about 21 MWd/kg, a 2-bundle shift fuelling scheme would be appropriate. optimae Th l fuel-management strategy depend sucn so h consideration fuee th l s enrichmentsa d ,an This then yields the following value for the price of RU: CRU the number and location of adjuster rods. With 0.9% SEU (or RU), a regular 4-bundle shift fuelling scheme would result in good axial power profiles. 1/F2 [CU30F j• - CUF + i I (F F 6 8' ' - SRU) - SRU • ACswU - ACu02l With 1.2% SEU, a regular 2-bundle shift fuelling scheme would result in acceptable power meanine symbolTh e th f go gives i n below wels ,a illustrativs la e parametervaluee th r sfo s distributions eithe cora n ri e without adjuster rods Brac e (sucth nucleas eha A r generating station (Referenc uses severar wa d fo 1 thesf 1 elo e values; referenc statee9 enrichmentse sth , burnups in Canada), or in the peripheral channels outside of the adjuster rod region (Figure 8).(> 13 In this and tails assumed). situation powee channele th , th rf decreas woulo d d ,an en de epea alonon t k lengta e e gth th f ho channel. This would be particularly attractive in those reactors in which fuelling is in the direction pric naturaf eo l uranium ($50/k) gU of coolant flow (such as the CANDU 6 and Pickering reactors). The power then would peak at the channele th f o inle d , wherten coolane eth t enthalp lowestys i thid s,an would improv criticae eth l CUF£ price of UFfi conversion ($8/kg U) channel power (power at which the critical heat flux is first reached). Cswu: pric enrichmenf eo t (S110/SWU) In reactors with adjuster rods, a regular 2-bundle shift fuelling scheme would result in an axial Cuo2= fuea pricl UO fabricatio f eo n ($275/k) gU power distribution tha depresses i t vicinite adjustee th th n d i f yo r rods (Figure 9)/' Thi13 s would natural uranium requirement for 1 kg of enriched uranium (6.51) increase the peak bundle power in the reactor, and result in power boosting during refuelling when separativ f enricheo g ek wor1 dr kuraniufo m from natural uranium feed the fuel is at relatively high bumups, which could lead to unacceptable fuel performance. Thus, (4.31 SWU) with enrichments of 1.2% or greater, other fuel-management options must be employed. Two options that have been studied are axial shuffling and the checkerboard fuelling scheme. premium for enrichment of RU($10/5WU) premium for UFe conversion using RU ($2 I/kg) 3.1.1 Axial Shuffling AC|jo2: premiu fabricatioa UO r mfo n fro (30/kgU mR ) Axial shuffling provides the greatest flexibility in shaping the axial power distribution. Axial - p2' recovered uranium requiremen enrichef o g k 1 d r uraniufo t m (5.03) shuffling involves removing some or all of the bundles from the channel, rearranging the bundles N) in pairs reinsertind ,an g fuel-bundlsome th f eo e pairs back int channee oth differena n li t order, -J separativ f enricheo g ek wor1 dr kuraniufo m fro(3.9U mR 5 SWU) along with fresh fuel. Axial shuffling is the reference fuel-management strategy for the newest ro 800 membe CANDe th f ro U family--the CANDU 3-sinc l fuellinee al th f dons o gi d een froe mon 00 reactor (i.e., uni-directional fuelling). Preliminary studies indicate that axial shuffling is feasible in other CANDU reactors as well. The ease of application of axial shuffling, and the number of bundles that can be removed from the channel, depend on several factors. For instance, the fuelling machine magazine capacity must be sufficien bundlee th f holo o t tl s d dischargeal d fro channele mth frese ,th h bundlee b so t tha e tar inserted into the channel, and other components, such as the channel closure and shield plug. The coolant flow must be sufficient to discharge all of the bundles. One axial shuffling fuelling scheme has been identified that is attractive for 1.2% SEU/14) This shuffling scheme is bi-directional, although in a reactor in which the fuelling is from one face of the reactor CANDe th , sucn i s Uh, bi-directionaa 3 l fuellin simulatede b n gca bundlf I . e positions along the channel are numbered from 1 to 12 starting at the "refuelling" end of the channel, then new bundle firse sar t loaded into position . Upo2 d nan subsequens1 t fuelling operationse ,th bundle shiftee positionn ar s i 2 followss da d an s1 : 2 1 1 1 0 1 9 8 7 6 5 4 3 2 1 1 -» 5 -» 3 -» 9 -» 7 -> 11 -» discharged Axial Bundle Position 2-» 6-* 4 -> 10 -> 8-» 12-» discharged Figur: e8 This axial shuffling scheme would be restricted to the central channels, in the vicinity of the Axial Power Profile For Regular 2-bundle adjuste rperipherae rodsth n I . l channels regula,a r 2-bundle shift fuelling scheme woule db Shift, Adjuster Rodt sOu employed. This axial shuffling scheme results in low peak bundle powers. Figure 10 shows a typical distribution of bundle powers along a channel in the vicinity of the adjuster rods with axial shuffling. Moreover bundlee ,th declinin a e sse g power history with increasing bumup, whichs i desirabl goor efo d fuel performanc t extendeea d bumup (Figure 11). 800 - 800 600 -...... I ^ 400 - ^^ 200 - 1 2 3 4 5 6 7 8 9 10 11 12 Axial Bundle Position 45672 1 1 1 0 81 9 Figur: e9 Axial Bundle Position Axial Power Profil Regular eFo r 2-bundle Figure 10: Shift, Adjuster Corsn I e Axial Power Profile For Axial Shuffling Channel K-11 (5543 kW) o o o o o o ooooooooooo o 5 Axial Bündle Position oooooooooooo o o 600. •——•n 3 „ ? Bündle2 ooooooooooooo o o o Ù-——0 Bündle 1 600 !i ———6 [ oooooooo«o«ooo o o o o oooooooo«o«ooo oo o o 5 S 1 4 oooo»o«o«o»o«o« o o o o o Ä 500. g £, 500- 1Q_ 1 — . — oooo»o»o«o«o«o» o o o o o o> 0~?"0 3.1.2 Checkerboard fuelling 800 Channel L-10, (6500 kW) checkerboare Th d fuel-management schem anothes ei r strategy that coul usee db accommodatdo t e 1.2% SEU in the vicinity of the adjuster rods/) 15 In the checkerboard fuelling scheme, adjacent channel pairs are refuelled vrilh different bundle shifts. Figure 12 is an example of a 6/2 checkerboard with 1.2% SEU. A 6-bundle shift fuelling scheme throughout the core would result in an axial power shape that is strongly peaked in the centre of the channel. A 2-bundle shift fuelling scheme peaks towards one end of the channel and is depressed in the centre of the channel if adjuster rods are nearby. The resultant axial bundle power distribution with the checkerboard is shapeso flattes i tw d blene a r,th an axiallf do y (Figure 13)6-bundle .Th e shift channe realls li ya pseudo 6-bundle shift, comprising exampler fo , , three quick 2-bundle shift successionn si n i , order to reduce the power peaking that would result from a 6-bundle shift. The power histories resulting from checkerboard fuelling should be acceptable from the perspective of fael performance. 3.1.3 Future Reactors Optimized for the Use of SEU 2 1 1 1 0 1 9 8 7 6 5 4 3 2 1 Axial Bundle Position In future reactors, many of the complications in fuel management resulting from the central location adjustee th f o overcomre b rod n sca repositioniny eb adjustere gth othed san r reactivity devicesA . Figur: e13 configuratio reactivitf no y device bees sha n identified that woul near-optimue db botr mfo h 1.2% to SEU and natural-uranium fuel.(>!6 This would allow the use of a regular 2-bundle shift with 1.2% Axial Power Profile For Checkerboard CO SEU, and a regular 8-bundle shift fuelling scheme with natural uranium. (Pseudo-6-Bundle-Shift Channel) CO 3.1.4 Strategie Introducinr sfo U gSE Ontario Hydro has irradiated more than 600 000 bundles containing UC>2 in its CANDU reactors, o with a very low bundle defect rate of about 0.1%. Over 3000 CANDU bundles have been fuel-managemene Th t considerations above focuse equilibriun do m fuelling. Another irradiate above-averago dt e bumups maximua o t , MWd/kgwitw 0 3 fe mf ha o . Abou0 15 t introduces i U reactorconsideratioe SE th unlikele s i n d i th t I . w yho woulthas ne i ton d starta bundles have experienced bumups above 17 MWd/kg, mainly in Ontario Hydro reactors, but with reactor wit fresha h cor f 1.2e o reactivit% e th SEU: y needin hele b do gt down woul vere db y relativel powersw ylo declinind ,an g power histories numbee Th . bundlef ro s drops sharple th s ya large, which would resul larga n i te los neutrof so n economy could ,an d pose problem safetn si y burnup increases. The experimental irradiations in research reactors associated with the initial and licensing. A more likely scenario would be to start with a fresh core of natural uranium, or developmen CANDf o t U fue involvel al l d enriched uranium. Som bundle6 e6 s have been with enrichment slightly above natural, and gradually introduce the higher enrichment. The irradiated to high bumup (maximum 45 MWd/kg) at high power in experimental reactors, transition from natural uranium fuel to 1.2% SEU has been studied for two cases: the supplemented by data from the irradiation of 173 single elements. checkerboard fuelling scheme with 1.2% SEU in a standard CANDU 6 reactor, and a regular 2- bundle shift fuelling schem a cor en i e wit witU h SE h1.2 repositione% d reactivity devices/17) The available data show thacurrene tth t 37-element fuel desig capabls ni reliablf eo e operatioa o nt bumup of 17 MWd/kg with a normal declining power history. Data also suggest that enhanced Both strategies were workable. fission-gas release, producing potentially life-limiting effects begiy occuma ,no t bumupt ra n si excess of about 19 MWd/kg. This effect is attributed to changes in fuel chemistry discussed 3.1.5 Reactivity Control earlier. Recently, post-irradiation examinatio severaf no (PIE) l fuel bundles that reached extended bumups The use of enrichment in CANDU will lower the reactivity worth of the adjuster rods (since the 18 enriched fuel is blacker than natural uranium). Fortunately, the reactivity worth required to provide in the Bruce power reactor was performed/ ) No evidence of fuel-performance deterioration was observeMWd/kg9 1 o t p du . However bundleo ,tw s havin averagn ga e discharge burnu6 2 f po a given xenon override time is lowered correspondingly, so the same adjuster rods provide about MWd/k havind gan g experience dpeaa k outer-elemen severad t poweha kW/2 5 ml f defectivro e the same xenon override time for both natural uranium and SEU. elements. Hig releas s ceramographa h ga measure s UO e wa 28%) o t . p d(u y revealed substantial The reactivity worth of the shutoffrods and zone controllers is also lowered, but is adequate. grain growth, and tear-out of grains at the fuel periphery, associated with extensive porosity. The Software changes to the control system parameters (such as modifications to the controller gains) oxide thickness on the outer sheath surfaces was up to 7 urn. Outer-element diameters showed 2% may be required. increase. Existing fuel codes under-calculated grain growth, fission-gas release and diametral increase. Preliminary examinations of three other bundles having lower bumups (average 21 MWd/kg) but higher linear heat ratings (maximum of 58 kW/m) revealed failures associated with 2 3. Fuel Performance Aspects high fissio releass n ga excessiv d ean e diametral strains/18) 3.2.1 CANDU Fuel Performance at Extended Bumup At extended bumups evidenc, w therno s ei e tha fission-gae tth s releas highee b n erca than expected from low-bumup extrapolations fror ,o m codes develope fuelbumupR w ,lo LW r fission dn fo I . - With the optimum SEU enrichment of 1.2%, a bundle-average bumup of about 21 MWd/kg is gas releases larger than expected have been reported for both steady-power and transient attained, with bundle maxim arounf ao MWd/kg0 d3 . Thi thres i e time currene sth t bumur pfo conditions. natural uraniu mCANDUn i . Potentially life-limiting factors tha importane tar extender tfo d bumup conditions include®: There appears to be a bumup-, and possibly Hnear-power-threshold at about 20 MWd/kg, above which enhanced fission-gas release (up to 28%), grain growth and diametral strain (up to 2%) may • mechanical integrity, observede b , with defect stress-corrosioo t e sdu n cracking/ inference 1'Th ) tha s extendedei e tth - • deuterium pickup and sheath corrosion, burnup effects are associated with a degradation in fuel conductivity and/or fuel-to-sheath heat • fission-gas release enhancement resultin highen gi r internal pressure, transfer, resultin acceleratiogn i thermallf no y activated processes primare Th . y defect cause th n ei • enhanced fission-product concentration, increasing susceptibility to stress-corrosion extended burnup Bruce bundles mentioned above was stress-corrosion-cracking failure through the cracking (SCC), and sheatbraze th t eha heat-affected-zon inboare th f eo d end-cap weld graie Th .n growt fissiond han - dimensional stability (changes induced by solid and gaseous fission products). high-bumuo releass tw e ga th en i p bundles were typica fuef lo l with operating temperatures higher than expected fro powee mth bundlesro historietw e th . f Howeverso , these observations neeo dt The effect that is currently regarded as most important in potentially limiting fuel life at extended confirmee b examinatioy db statisticalla f no y significant sample. bumup is fuel chemistry. Several international programs are addressing fuel-chemistry effects (Halden, Studsvik, Riso, EPRI, BelgoNucleaire) inventore Th . heavf yo y metal, oxyged nan power-ramA p recentls teswa t y performe fuen do l fro Nucleae mth r Power Demonstration (NPD) fission products in UÛ2 is modified at extended burnup. Some of the oxygen present is fixed by reactor as part of the AECL extended burnup program/')1 The fuel had a bumup of 35 MWd/kg at the fission products, and some of the resultant compounds are more stable than if the excess a declining power history, and the NRU reactor at Chalk River was used to ramp this fuel to about oxygen was fixed by UO2-PuOa. The rare earths, Nb, Zr, Ba and Sr, form stable oxides, but 36 kW/m. The NRU ramp proved sufficient to cause significant fission-gas release and diametral fission products like Cs, Mo, I, and Te do not. At high temperature, the latter compounds are less strain in the fuel elements, and although no defects occurred, incipient cracks were detected in the stable than the fuel phase. In a closed system, oxygen is transferred from these fission products fuel sheaths. There is thus insufficient fuel performance data for the 37-element bundle to be towards the fuel phase; above some temperature the fuel phase then becomes hyper-stoichiometric. assured of acceptable performance at extended burnup, particularly with load-following. deviatioa f I n from stoichiometry takes dominanplace th n ei t fuel phase, significant changes occur in thermal properties (thermal conductivity), diffusion propertie mechanicar so l properties (creep Technical improvements could be made to the internal design of the 37-element bundle to meet the rate). Such a deviation will significantly influence fuel-rod behaviour, characterized by higher performance challenges of extended bumup, such as improved sheath coatings, zirconium-barrier temperatures and increased fission-product release. sheath, optimized pellet configuration and density, or graphite-discs. analyzinn I g extended-bumup performance requirements assessmene th , thas wa t most long-term The CANFLEX Development Program has been underway at the Chalk River Laboratories of benefits for CANDU would be achieved with a new bundle design. A new bundle design with AECL Research since 1986 February. It comprises performance testing, manufacturing, wide application in current and future CANDUs, for natural uranium or advanced fuel cycles, thermalhydraulics confirmation, reactor physics and endurance/handling tests. An important coul provee db n wit same hth e developmen demonstratiod tan n effort require modifiea r dfo - d37 componen CANFLEe th f to X produc date th a s bas ti modellin d ean g capability generated during element design. development, to support large-scale manufacture and future licensing. The CANFLEX advanced fuel bundl wids eha e applicatio currenn i futurd tan e CANDUs optimae th s i d l carrie,an r rfo advanced fuel cycles. 3.2.2 The CANFLEX Bundle«, 20) The new CANFLEX bundle has 43 elements, and features two element sizes. CANFLEX is a logical extensio existinf no g technology: progressive CANDU fuel designs have increasee dth 3.3 Licensing Issues number of elements in a CANDU bundle from 7 to 19 to 28 to 37, and now to 43 (Figure 14). greatee Th r subdivisio alss ni linon i e with technology trend advancen si designsR dLW . AEC performes Lha d several safety studies fuelwitU h,SE both internall variour fo d ysan clients, Compared with the current design, CANFLEX will provide about a 20% reduction in peak linear and these have not identified any safety problems. In terms of releases during potential accidents heat ratings, which will enhanc bumus eit p capability bundle Th . compatibls ei e with existind gan (suc loss-of-coolant)s ha highee th , r absolute inventor fissiof yo n products (du increaseeo t d future (for example, CANDU 3) fuel-handling systems. Some key points showing the wide bumup) is balanced by a smaller fractional release (due to lower element ratings with the applicatio CANFLEf no X are: CANFLEX fuel bundle, and fuel-management schemes featuring a declining power history with bumup). in current and future reactors, lower element ratings at current bundle power, in future reactors, power uprating without exceeding current element ratings, achievemen extendef to d bumup facilitate optioy db lowef no r element ratingd ,an optimizatio internaf no l design, natura enricher lo d fuel, 4. IMPACT OF EXTENDED BURNUP IN CANDU ON THE FRONT END OF increased operatin safetd gan y marginsd ,an THE FUEL CYCLE power manoeuvring facilitated. 4.1 SEU CANDU reactors currentl naturae yus woull U uraniuSE d f requiro me UÛ2us e additionaTh . l productioe stepth n si CANDf no U fuel: conversio UsOf no UFego t , enrichment, conversiof no UFft to UO2, followed by fabrication of enriched UÛ2 fuel. Conversion of UjOg to UFe is provide CAMECy db CanadaOn i facilitA . y woul requiree db producconvere o dt UF e ttth from the enrichment plant to UO2- While this has been done on a small scale in the past in Canada, a new facility would likel requiree yb handlo dt e large volumes. While enrichment servicee sar available commerciall worlde th n yi , CAMEC Ocurrentls i y assessin feasibilite gth CRISLe th f yo A (Chemical Reactio Isotopy nb e Selective Laser Activation) enrichment process^ availabilite 2')Th . y of a domestic enricher in Canada would make the use of SEU even more attractive, and should be viewe industrian a s da l opportunit Canadar yfo . fuee th l t fabricatioA n plant, changes woul requiree db manufacturo dt e CANDU bundles from SEU (for example, to ensure criticality limits). Since enriched fuel is routinely fabricated by PWR fuel fabricators, this is not seen to be a problem. Douglas The use of the CANFLEX bundle for SEU will also impact on fuel fabrication. There are two 19-elemenf element sizes, compared to one in current CANDU bundle designs. However, the number of distinct element types (including number and orientation of bearing pads and spacer pads) may be simila 37-elemene th o rt t bundle, dependin finae th l n gdesigno fabricatioe Th . - 28 ne costth r sfo elemen 37-elemend tan t bundle similare woule sar on o d,s expect simila CANFLEre costth r sfo X bundle in equilibrium, although there will be initial investment costs to tool-up for a new bundle NPD design. Fuel cycle calculation conservativelU SE r sfo y assum cos% t 20 penalt ea CANFLEr yfo X 7-element CANFLEX fabrication. 43-element extendef o e Whilus e de th bumup will reduce quantite th CANDf yo U fuel bundles required from fabricator lone th gn stermi anticipates i t i , d that this woul compensatee db increasey db d nuclear Figur : CANDe14 U Bundle Evolution demand. CO 4.2 . 6 CONCLUSIONS (VI radiologicae Th l implication fabricatinf so g CANDU fuel frocontaininU mR g 0.8-0.9% U-235 There are resource and economic incentives for extending fuel burnup in CANDU through the use have been examined.' importann 22A ) t conclusio thas fabricatioe ntth wa highlf no y purifieU dR . WitRU hr SEUo U , SE natura f o l uranium requirement reducee sar 20-40%y db fuellind ,an g present mor o hazarsa n f eo d from interna externar lo l exposures than doe fabricatioe sth fresf no h costs by 20-35% relative to natural uranium fuelling, depending on the enrichment level and LWRfuel. Although Cs-13 Sr-9d 7fission-producan e 0ar t particulaimpuritieo n , RU n rsi assumptions. The optimum enrichment is about 1.2%. The use of RU in CANDU is very difficulties are foreseen in CANDU fuel fabrication, as long as their concentrations are low. attractive: doubl energextractee e eb th n CANDyn ca i burnin y db U R e Ugth compare - re o dt PWRa enrichmen potentian e i e Th . us d l tan saving fuellinn si g cost substantiale s ar wit U hR . AEC currentlLs i y engage cooperativa n di e program with COGEM Francf Ao asseso et e sth fabricability of CANDU fuel from RU. Of particular interest is whether Cs-137 would be driven The on-line metho fuellinf do g CANDU provides both opportunitie challenged san r sfo off during sintering, and condense in the colder parts of the furnace, resulting in increasing fields accommodating enriched fuel. The opportunity is in the flexibility of using a variety of fuel over time experimentae .Th l program will confirm tha concerna t t thino s i . management strategies challenge ;th designinn i s ei gfuea l management scheme that ensuresa declining power history with bumup. With 0.9% SEU (or RU), a simple 4-bundle shift fuelling schem acceptables ei . With 1.2% SEU, more imaginative fuel management scheme availablee sar , 5. IMPACT OF EXTENDED BURNUP ON BACK END OF THE FUEL suc axias ha l shufflin checkerboardr go . CYCLE At extended bumups, there is now evidence that fission-gas release can be higher than expected Increasin bumue gth CANDf po threr o Uo e factoa fue woultw y f b l o r d reduc discharge eth e rate from low-bumup extrapolations. There also appears to be a burnup-, and possibly lincar-power- of irradiated fuel from the reactor by the same factor. This significant reduction in the volume of threshol about da MWd/kg0 t2 , above which enhanced fission-gas release, grain growtd han spent fuel produced could be a positive step in increasing public acceptance of nuclear power. diametral strain may occur, leading to defects due to stress-corrosion cracking. The CANFLEX bundle is being developed to provide assurance of good fuel performance at extended bumup in There are also economic incentives from the back-end of the fuel cycle to use SEU fuel in CANDU. CANDU. Figur show5 e1 s that disposal costs (per unit energy reducee )ar d significantly with geologicae th SEUr Fo . l disposal concept develope AECLy db , with spent natural-uranium fuel, the separation of the bore-holes in the disposal vault is limited by the physical properties of the References' rock. With SEU separatioe th , bore-holee th f no limite heas si e th t y generatedb usee th dy dfuelb . Hence, there is an economic incentive (from the perspective of disposal) for longer cooling periods . 1 P.G. BOCZARet al., "Slightly Enriched Uraniu CANDUmn i Economin A : c First Step with SEU (such as 50 years). This has to be balanced with increased storage costs. Towards Advanced Fuel Cycles", AECL-9831 (1988). 2. IJ. HASTINGS, P.G. BOCZAR, and A.D. LANE, "CANFLEX: An Advanced Bundle Desig Introducinr nfo g Slightly Enriched Uranium into CANDU", AECL-9766 (1988). 0.8 •\ i 3. P.R. BURROUGHS and G.H. ARCHTNOFF, "Slightly Enriched Fuel for CANDU Natural Uranium Reactors", Proc. 27th Annual Conference of the Canadian Nuclear Association, Saint John, 1987 June 14-17. i c » p . 4 R.E. GREEN, P.G. BOCZAR . HASTINGSIJ d an , , "Advanced Fuel Cycler sfo CANDU Reactors", AECL-9755 (1988). 8 5. R.W. MORRISON, "Uranium Supply: Canada in the Context of the World Situation", o Proc. 28th Annual Conferenc Canadiae th f eo n Nuclear Association, Winnipeg, 1988 June "tô 12-15. m . . 1.5U %SE o » i . J.G6 d an .H.K MELVINE .RA , "Uranium Enrichment Opportunitn A : y Window", ibid. h o 8- p . 7 "The Economic Nucleae th f so r Fuel Cycle", OECD Nuclear Energy Agency, Organisation for Economic Co-Operation and Development, 1985. 10 20 30 . 8 "Economic CANDU-PHW"f so , Ontario Hydro Report, NGD-10, 1985. Burnup (MWd/kg) 9. IJ. HASTINGS, P.G. BOCZAR, C.J. ALLAN and M. GACESA, "Synergistic CANDU- LWR Fuel Cycles", AECL-10390 (1991). Figure 15: Effect Of Burnup On Disposal Cost AECL-XXX XAtomis i c Energy of Canada Limited published report. . 10 P.G. BOCZAR . HASTINGJ . I , CELLI. A d San , "Recyclin CANDn gi Uraniuf Uo m NECESSITY FOR AND FEASIBILITY OF and/or Plutonium from Fuel"SpenR LW t, AECL-10018 (1989). BURNUP EXTENSION IN CHINA'S 11. "Plutonium Fuel, An Assessment", A report by the Nuclear Energy Agency (NEA) of the NUCLEAR INDUSTRY Organisation for Economic Co-operation and Development (OECD), 1989. . 12 J.B. SLATE GRIFFITHS. J d Ran , "Physic Economid san c Aspect Recycle th f so f eo Zhenglun LID LWR Uranium in LWR and CANDU", Proc. 5th Annual Canadian Nuclear Society Shanghai Nuclear Engineering Research and Design Institute, Conference, Saskatoon, Saskatchewan, 1984 June 4-5. Shanghai, China 13. M.H. YOUN1S and P.G. BOCZAR, "Equilibrium Fuel-Management Simulations for CANDa n i U 1.2, AECL-998U%6" SE 6 (1989). Abstract 14. M.H. YOUNIS and P.G. BOCZAR, "Axial Shuffling Fuel Management Schemes for The Incentives to extend burmip are present in China, even the 1.2% SEU in CANDU", AECL-10055 (1989). nuclear power development is at the starting stage. The studies for extending burnup is based on that the civilian nuclear power is the 15. P.S.W. CHAN and A.R. DASTUR, "Checkerboard Fuelling, the Key to Advanced Fuel priorit nucleae th f yo r industr d thay an tfair-size a d fuel Cycles in Existing CANDU Reactors", Proc. Sixth Annual Conf. Canadian Nuclear industrial comple bees xha n forme Chinan i d . Society, Ottawa, Canada, 1985, June. The Qinshan NPPs' reactor design and in-core fuel management are . 16 P.G. BOCZA "ImproveK M.TDY d n Ran d.va Locations of Reactivity Device Futursn i e briefly presented case Qinshaf th eo n I . n HPP-1 fuee ,th l cycle CANDU Reactors Fuelled with Natural Uraniu Enrichemr o d Fuels", AECL-9194 (1987). economic evaluation using the common economic input parameters are conducted e studTh . y results show that extension burnu vers pi y . 17 P.G. BOCZAR, N.B.Y. CHENG J.Wd ,an . THOMPSON, "The Transition from Natural beneficia reducinr fo l fuee coste th e l th g f th scyclo d ean Uraniu CANDa n i m 1.2o t U %USE with Repositionin Reactivitf go y Devices", Second utilizatio nucleaf no r resource. Int. Conf. CANDU Fuel, Pembroke, Ontario, 1989 October 1-5. A Graduate Research and Development Programme for extended burnup is . NOVAJ I.J. d K.an 18 HASTINGS, "Ontario Hydro Experience with Extended Burnup, taking place in China. Power Reactor Fuel", AECL-10388 (1991). 19. TJ. CARTER et al., "Effects of End-of-Life Power Ramping on UO2 Fuel", AECL-10417 (1991). . 20 I.J. HASTINGS, A.D. LAN P.Gd Ean . BOCZAR, "CANFLE Advancen A X - d Fuel I. INTRODUCTION Bundl CANDU"r efo , AECL-9929 (1989). e NucleaTh r Powe s beeha r n playing importan e woldwidt th rol n i e e 21. Nuclear Fuel, 1991 August 20. energy electricity generation. . CELLA C.Wd I an . . TURNER22 , "The Radiological Implication Fabricatinf so g CANDU The safety and econony are the najor issues for the Fuel From Recycled Enriched Uranium", AECL-4050 (1987). development of nuclear power. One of the important areas is the effective utilization of nuclear resource and reducing nuclear power costs. e burnuTh p extensio n interestina s i n g studye subjecth n i t nuclear industry of China as around the world, even the nuclear power developaent is still at the starting stage. The study is based on that the civilian nuclear power is the priority of the nuclear industry and that a fair-sized fuel industrial conplex has been formed in China. The technolog r PWfo y Rs e i adopte coa»erciath r fo d l nuclear 00 03 power plants. GJ The Qinshan projec a Chines s i t e designe d constructean d d PV/R w specifiThlo e c ratings ,it LHG thereforew s i lo R e th , -t». NPP [1] with'300 We capacity and a design burnup 24.0 KVId/KgU obvious characteristic. The core design is a comprehensive for early stage. result of «any factors, since the core design involves trade-offs between the nuclear, theraal hydraulic and mechanical The project will contribut o t gainine g experiencd an e requirements e conservativth o S . e design philosoph s i adoptedy . •astering the technology for further development of nuclear power The low LHGR design is not only beneficial to the safety and n Chinai . operational flexibility but also to fuel manageaent, it offers plenty of rooa for various loading patterns, being able to e y projecDay Th n importeBa a a s i t d PW witP RNP h e 2x90MW 8 accomplish multi-region loadin d multi-batcan g h refueling. capacity and a warranted design-discharge burnup of 33 MWe/KgU. There are possibility and necessity of extension burnup for both NPPs, for the benefit of econoay and technology. 2.2. In-core Fuel Management During the process of reactor design and nuclear fuel Research and Development, some works on fuel performence, in-core fuel For the history reason, a rather low discharged burnup was management as well as economic evaluation of fuel cycle cost have designed. After gaining operation experiences, higher burnup been already done. shoul e achievedb d . A high burnup programme is in place in China. The nuclea e s i Qinshath designe r P r d fueNP fo an n dl •anufacture y Chinesb d a efforvasn d ow et an tamoun f o Researct h and Development have been conducted (3] e fueTh . l in-pe il testin s beegha n completed [4], . 2 QINSHAN NPP'S REACTOR DESIG FUED AN NL MANAGEMENT The systematical analysis nodels and corresponding computer programs have been developed for fuel design, core design and in- Z. \ .Reacto r Nuclear Design core fuel nanagement. The Qinshan NPP belongs to the snail size PWR plant, but its The work ) whic(5 s h were already e finisheQinshan'th r fo d s features are nearly the same as the current PViR type plants in fuel nanagement are- e worldth ] [2 . - Refueling calculation for cycles 1 to 8 with burnup The aain characteristics of Qinshan Reactor are listed as analysis, in the mode of Out-In-In follows: -Study for extending the burnup utilizing the moderator's Electrical output, MYfe 300 tesperature effect d fuel'an s s Doppler L effectEO t a s Thermal outputW M , 966 Core activity heightn i , 2.9 -Analysis of the effects of increasing the equilibrium Equivalent core diametern , 2,486 refuelinq U-235 enrichment Number of fuel assemblies 121 Rod array 15x15 - Improving burnable poisons using B.Silicate Glass instead Fuel weight (U)2 t U0 , , 40.146 (35.917) of B.S.S. Wate o fuet r l volune ratio 2.065 Average LHGR, W/cm 135 - Fuel management Leakage(LLanalysiw Lo r w fo d sVer Lo an ) y Max i nun LHGR, W/cn 407 Leakage(VLL) e ,mode th bot f In-Out-In o i sh d In-In-Ouan n t Fuel enrichment (first core), w/o 2.04; 2.67; 3.00 Fuel enrichment (refueling), w/o 4 3. 3. - 0 - Low parasite design of the structural materials Equilibriua cycle burnup, MWdxKgU 0 3 2- 4 - Optimization of in-core fuel management 3. ECONOMIC EVALUATION FOR FUEL CYCLE COSTS For a given cycle energy generation (e.g. a 12M cycle and 300 MWe output), as the discharge burnup increases, the equilibrium There are some analysis models and codes for fuel cycle batch size decreases, the reload enrichment increases and the econonic have been develope y varioub d s institution. ] [7 ] [6 s average in-core residence time increases e trendTh n .i fues l Among these models, the sane principle - levelized discount cost cycle cost then becom a etrade-of f amon e opposinth g g trendf o s •odeI, is used. the various cost components. e basiTh c factors which coste i»pac th f fue o sn o tl cycle are: The direct cost, generally, is decreased with the extended discharge burnup, provided the reload enrichment is not too high. basie th c- pric f Material,o e ' processin d servicean g s However e indirecth , t costs tren usualls i d y opposite because - economic parameters, and the effect of the energy discount factor is unfavourable when the discharge burnu s increasedi p . - technical performance parameters. In addition to sensitivity study for the basic price of According to the »ode I and data recommended by IAEA, some various material and processing, the economical and technical evaluation e fueth l r cyclfo s e cost e beinar s g conducted. analysis, are necessary. The analysis node I used was checked by the benchnark, and the Fro« the technology's point of view, the study should results consist with IHAE's result. emphasiz hige th h n o burnue p fuel performanc e feasiblitth d an e y e fueoth fl managemen r sucfo thiga h h discharge burnup. For example, in the case of Qinshan NPP-1, the evaluation results adopting common economic input parameters are as follows: 2 Q T Bu C5 Cost 4. A HIGH BURNUP RESEARCH AND DEVELOPMENT PROGRAMME (KgU) (Years) (MWd/KgU) (w/o) (Mills/Kwk) The study results of other countries and China show that the 121/40 11874 3.03 23.43 2.90 9.104 extension burnup is very beneficial for reducing the fuel cycle's 121/30 8906 4.83 31.24 3.30 7.640 costs and the utilization of nuclear resouce. 121/20 5937 6.05 46.86 4.35 6.693 121/15 4453 8.07 62.48 5.40 6.325 According to the status and development requirement of nuclear industr Chinan i y Researca , d Developemnan h t programm n o hige h burnup is proceeded. where Z - ratios of core load fuel asseablies to reload fuel assemblies; A 3-phases Research and Development plan was suggested 18}. Q - batch U invetory, KgU; The objectives and developing stages are as follows: - fue T l residence in-core time, years; Stage Objective Year - Discharg u B e burnu r equilibriufo p m cycle, MYfd/KgU; Phase-1 25 - 33 MWd/KgU 1996 Phase-2 MWd/Kg5 4 4- 0 U - 2000 C5 - reload enrichment, w/o Phase-3 50-60 MWd/KgU - 2010 n I this study e technicath , l parameter e takear s n from The Research and Development on extension burnup consists of CO Qinshan-1 design value as above, and the common econnic input en parameters of WREBUS report is adopted. basic research, specific subject d comprehensivan s e studies. G3 The basic research includes: Nevertheless, the most valuable issue of the programme is to O> accumulate practical experiences from the Qinshan and Daya Bay - économie analysi f fueo s l cycle costs NPPs operation examino t s e fuea s th el performanc o t chec d kan e the analysis models. - fuel performance data bank establishment 5. CONCLUSION - computer packages development and improvement For development of civilian nuclear power programme, - set up the standard and codes obviously, mastering the technology of extension burnup is not only necessar t alsbu yo possibl Chinan i e . -measurement, inspection techniques and facilities development A gradual Researc d Developmenan h t programm higr fo eh burnup will be accomplished by our own efforts, and the international e specifiTh c subjects consisi of t cooperation is also needed. - developin parasitw lo g e design Since the international technology exchange and cooperation supported by"IAEA is very successful and beneficial, the related developin- g Gd203-U02 fuel activities shoul e continuelb d y conducted. - waterside corrosion - fuel in-pile testing REFERENCES The comprehensive studies consist of: I' (luyang Vu. [lcsi<)ii ;unl Con s t ruf I inn of Ijinshan unclear powr ptiitit h PacifiVt . c Banks Nuclear Conference 1985y Ma ,. - testing of fuel assembly performance [2] Liu Zhenglun, An Overview on Reactor Core Design and Analysis - themat-hydraulic testing Method for Qtnshan NPP. GRS-SNEROI Seminar on "Reactor Core - Mechanical testing Design Sept. 12-16. 1988. Shanghai. - compatiblity testing u Slienghiia[3T ] , etr. :)QO MR FueWPi l Assembly Design, Testing - lead fuel testing in reactor and In-p f I i test. Chinese Journa f o Nucleal r Sciencd an e Engineering, Vol. 9 No. I. Mar. 1989 - PIE [4] Song Vienlmi, Analysis and Evaluation of the Fuel Element for - advanced fuel nanagement : optimalization of refueling and Qinshan NPP, GNIC-OIUO. If39. reactivity control ; 5 : Clidi Pingdnif), Tin1 In-core Fuel Managemen d Burmi[an t i - coast-down operation Analy-i r ginshafo ? n NIT. ;:\EKDI Internal Report. I1)!!1; Among*the subjects mentioned before e mosth t, important ones ;6 Liu Zhengliiii. etr. A Common Model for FuH Cycle Cost are; Analysis. SNF.RUI Internal Report. Ifltï. - advanced burnable poison Gd203 - U02 n Dingqin[7Li ] . Economic Benefit f o Increases d Discharged Biirmi)) r PVtR,fo . CNIC-DÖ549. 1991. - waterside corrosion >8] Zltanq Zhongyue « ChenqxinLi . , Som e eResearc th Idea r d fo s an h - advanced fuel management Development Progra f Higmo h Burnup Fuel. (Wil e published)b l . PRESENT STATUE TH F SO Power Plant Start-up Spent Pool NUCLEAR FUEL CYCLCSFE TH RN EI Fuel Capacity tU/a tu NPP Dukovany unit 1 WER 440/213 05.85 13.97 38.34 F. PAZDERA NPP Dukovany unit 2 WER 440/213 03.86 13.97 38.34 Nuclear Research Institute, NPP Dukovany unit 3 WER 440/213 12.86 13.97 38.34 Rez, Czechoslovakia NPP Dukovany unit 4 WER 440/213 07.87 13.97 38.34 NPP Mochovce unit 1 WER 440/213 12.93 13,97 71.9 NPP Mochovce unit 2 WER 440/213 09.94 13.97 71.9 Abstract NPP Mochovce unit 3 WER 440/213 09.95 13.97 71.9 NPP Mochovce unit 4 WER 440/213 04.96 13.97 71.9 The paper gives an overview of present fuel cycle activities in CSFR and NPP Temelin unit 1 WER 1000/320 05.94 23.36 273.05 perspectives at its development. Discussed are results of preliminary NPP Temelin unit 2 WER 1000/320 11.96 23.36 273.05 evaluation of alternative fuel cycle options. Main attention is devoted to technical and economical problems of extended burnup, with special emphasis All reactors under operation operate wit 3 yeah r cycles (some low-burnup on WER reactors. The most important part for further consideration is the fuel assemblie years)4 s cor n r i (FAreacto e efo e .Th ) ar fued ran l design e nucleabackenth f o rd fuel cycle, especialle once-througth r fo y h parameter n TABLi s 1 e assumei E[2,31 ar st I .d that reactors under alternative. construction will start with three year cycles. 1. INTRODUCTION TABLE 1 The IAEA Water Reactor Extended Burnup Study has indicated strong economic incentive r burnusfo p extensio s beeLWR'sn ni ha nt I .show n thae maith tn driving backen e forcth s ei d cost, alik whether efo r once-through cycld ean Parameter WER 440 WER 1000 closed cycle. e IAETh A International Conferenc e Safetth f Nuclean o yo e r Powe] 11 r discussed the problems of final disposal of high level waste, and concluded that burnup possibilite extensioon e reduco th t spens e w ni e th yho t fuel's Cladding outer diameterm ,m 9.1 9.1 amount. Also situation in long-term spent fuel storage could be improved Cladding minimum thicknessm ,m 0.6 0.63 with burnup extension. Fuel cladding diametral gap, mm 0.12-0.27 0.16-0.27 Gas plenum volume, cm 4 11 e nucleaTh r fuel w beincycl no n CSF i ges i Revaluate d from different Fuel rod length, mm 2570 3840 reasons, both economica strategicd lan . Becaus finao en l decisions wert eye Fuel column length, mm 2420 3510 reached, the paper gives only a review of present situation and the Minimum fuel density, g/cm 10.4 10.4 preliminary evaluation of some alternative. Central hole diameter, mm 1.2 1.4/2.4 Fill gas He He The technical problems connected with burnup extension were already Fill gas pressure, MPa 0.1/0.5 2.0-2.5 discussed withi e WREBUth n Se include ar repor d an td herr discussiofo e n Reactor thermal powerW ,M 1375 3000 purposes. Reactor electric power, MWe 440 1000 Coolant pressure, MPa 12.5 16.0 2. PRESENT SITUATION AND PERSPECTIVE Coolant temperature - inlet, C 268 290 Coolant temperature - outlet, °C 296 320 At present there are following NPPs in operation or under construction in Mean linear heat rating, kW/m 12.7 16.7 the CSFR [21: Maximum linear heat rating, kW/m 33 49 Average burnup, MWd/t 30000 28000/40000 UO- inventory in fuel rod, kg Power Plant Start-up Spent Pool 1.082 1.575 Fuel Capacity Number of fuel rods in assembly 126 312 tU/a tu Number of fuel assemblies in core 349 163 NPP Bohunice unit 1 WER 440/230 04.79 13..97 38.34 U inventory in assembly, kg 120.2 433.2 NPP Bohunice unit 2 WER 440/230 06..80 12. 53 38.34 U inventory in core, t 41.9 70.6 NPP CO Bohunice unit 3 WER 440/213 11..84 13..97 38.34 NPP Bohunice unit 4 WER 440/213 09..85 13..97 38.34 e front-enth w no dR o reactorr t otheoperationWE fo p l U s ral Ca sr fo sj ( Alternative Fu»l Cycl«* Comparison Alternativ» fuel Cycles Comparison 0 0 former COMECOM countries providee USSRe )ar th .y db Thi s monopoly position CSFR CSFR differencee ith s o mainlt e du sy between d WER'cord thosan an e A f sF o e Natural U Consomption Natura ConsumptioU l n other PWRs t presentA . , some diversificatio n thii n s respec s undei t r consideration. u t Nat. U . Nat. U. ktU e originaTh l negotiated condition e transporth f o s f speno t e t th fue o t l USSR after 3 years storage were 10 years ago modified to prolonged storage (about 6 years) which forced the decision to construct Intermediate Spent Fuel Storage Facilit spene th ye lac tf (IMSFSF)th o k fue o t le storag.Du e capacity before startuIMSFSFe th f .o pa certai s (aftewa n s numberFA f o r corresponding negotiations) transferred to the USSR. By the end of 80's the overall situatio s drasticallnha y change d lonan dg term storag r otheo e r alternative solutions are necessary. The present spent fuel storage capacity will be exhausted in 1994. Several options to solve this problem are under consideration mose th ,t probable being constructio f IMSFSno t Fa Dukovany spene .Th t fuel storage capacitie followinge sar : Facility Capacity Start-up -1000 tu 0 20*»1 0 o 209m 0 0 « 207»7 20 0 0 »7 1990 3010 2030 20SO 2070 2090 IMSFSF Bohunice 600 01.86 Year IMSFSF Dukovany ~ 2000 ~ 01.94 1 -"Go»*M C« CO« M3 •—Co« «i —COM 1 COM * ---c«*4 As the delay in planned start-up of IMSFSF Dukovany could have unfavorable consequences, a preliminary evaluation of several alternatives of the whole nuclear fuel cycle (under the light of new economical and political conditions performeds )wa . Four alternatives were assesse ] (Fig.1-6.)[4 d : once-throug- h fuel cycle with three year cycles, Alternativ« Fuel Cycles Comparison Alternative Fael Cycles Comparison once-throug- h fuel cycle with four year cycles, CSFR CSFR close- d cycle with early reprocessingd ,an Enrichment Services Enrichment Services - closed cycle with delayed reprocessing. EnrlchmenntU SW t . Enrlchmennt. kt SWU The preliminary conclusion are that extended burnup is the most direct way to reduce fuel cycle cost decisioe .Th choicf no e between once-througd han closed cycl s rathei e r speculativ s influencei d an e d mainl y threb y e important factors: 1000 - relatively well defined and high cost of reprocessing and Pu and U recycling, therefore taken in the time they arise, undefine- d cosspenr tfo t fuel fina wastA lR disposar e fo fina d lan l disposal, therefor somn i e e countries immediate political decision about financin spent a e t gfe fue a thi ly sb discharg e from reactor cors ewa considered, and clearlt no - y seen tim finaf eo l spent wast A R fue r leo disposal. Because of above mentioned factors and not easily chosen long term-value of discount rate, time schedule and financial problem issues make an actual 1970 »90 2010 2090 2050 2070 2090 1970 1990 2010 2030 2090 2070 2090 problem. Year Year Under above assumptions the ranking of alternatives is as follows' I —M ————CO 4 Cot K ** C« - - - C«4 « I —« "———Cm C«t »t Co*- - - Cai*4 ** 1 Closed cycle with delayed reprocessing 2. Qixce-through fuel cycle uith three year cycles 3. Closed cycle with early reprocessing Fig. Alternativ1 , e Fuel Cycle Compariso CSFr nfo R Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison CSFR CSFR CSFR CSFR Fuel Fabrication Cumulative Fuel Fabrication Spent Fuel Disposal Cumulative Spent Fuel Disposal Fabrication. ttU Spent FuelU t . Spent FuelU kt . 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2010 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 Year Year Year Year —— Cat *1 —— Cos «2 • " CatCo» *4 *3 —— Cot* I —— Cat« 2 Cot* * - • • Cai* 4 •—— Cot* 1 —— Cat* 1 Cat« > - Cat* 4 —— Cot *I —— Cot *2 Cot* - Cot-• *4 ) Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison CSFR CSFR CSFR CSFR Intermediate Storage Capacity Fuel Reprocessing RA Waste Disposal Cumulative RA Waste Disposal Spent Fuel. ktU Spent FuelU 1 . Spent Fuel, tu Spent FuelU kt . 300 » « / 250 5 200 200 1 i 4 i 150 150 ' ' '; • ' • i th •.,-•' 3 too * "—\ 1 2 50 i 1 - n i . n> , i 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2060 2070 2090 Year Year Year Year "——Cat* 1 ——Co*« 3 Cotm 3 Cat* 4 —— Cot* 1 —— Cat« 2 Cot* 1 * - Cat« 4 Cot t * Cot »» Cot*4 C«l «I —— Cot *2 • Cot• Cat« 4 *4 CO CO Fig. 2. Alternative Fuel Cycle Comparison for CSFR Fig. 3. Alternative Fuel Cycle Comparison for CSFH -p.. Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison o CSFR CSFR CSFR CSFR Annual MOX Utilization Cumulative MOX Utilization Annual Fuel Cycle Cost Cumulative Fuel Cycle Cost Fuel, tU Fuel. IU Cost. MS Cost. GS 70 60 600 1500 SO 500 I " 40 400 - 0 3 100 10 200 10 100 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 Year Year Year Year —— Co«« 1 ——COM] Co»« 1 ---Co*»« —— Co»« I — Co»« 2 Cot* 3 • Co»« 4 —— COM I ——Cat« 1 Col« 3 ---Co»« 4 —— Co* «I —— Co» «2 - - ColCO»« 4 « 3 Alternative Faol Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison CSFR CSFR CSFR CSFR Annual Recycle UtilizatiodU n Cumulative Recycle UtilizatiodU n Annual Fuel Cycle Cost Cumulative Fuel Cycle Cost Fuel. tU FuelU t . Cost (discounteS M , 1991o dt ) Cost (discounte$ G . 1991o dt } ouuu 2600 t ! i 2000 i il ^ s 1500 ,' \ : i / 1000 I \ ' / 500 ' n 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 1970 1990 2010 2030 2050 2070 2090 Year Year Year Year —— Co» I * ——Co« »2 Co» «3 ---cai«4 —— Cot« I ——COB« 2 Co»« 3 -"Co»« 4 —— Co» «I —— Cot «1 • ColCat «3 «4 —— Caw i -— Co«* J Coi« 3 " Ca»« * Fig. 4. Alternative Fuel Cycle Comparison for CSFR . Fig5 Alternativ. e Fuel Cycle Compariso CSF"r nfo R Alternative Fuel Cycles Comparison Alternative Fuel Cycles Comparison 3.1. FROND TEN CSFR CSFR RA Waste Storage Funds Availabl r Repositoreto y Majority of the front-end activities are not covered in Czechoslovakia, nevertheless the economic impact of possible technical problems is also not Spent f net. ktU ______Funds. MS (discounted to 1991) without importance. The main technical problems could be: - increase of required enrichment, fabricatio- higf no h enriched, high burnup fuel assemblies: - enrichment above licensed enrichment value, - application of burnable absorbers, fabricatio- fuelX MO ,f no enrichmen- recyclef to , dU - fabrication of FAs with recycled U. Specific problemR reactorWE f o s s coul e connecteb d d witw spacene h r grids. Of interest is also fuel recycling in other advanced type reactors (e.g. CANDU) 1970 IWO 2010 3030 2050 2070 2090 1770 IVVO 201 » 20 2090 0 2070 2090 Year Year —— Caul ——Caul Co«» ---Co»« —— Cau4 l« 3 C« —u —" - Ca Cttl «a Fig . 6 .Alternativ e Fuel Cycle Compariso CSFr fo nR 3.2. DESIGN AND OPERATION maie Th n possible problem e solveb o connection t si d n with extended burnup were discussed withi HREBUe th n S Study [3]: - design of extended burnup FAs, - testing of lead FAs, - reactivity coefficients More detailed analyses are however necessary. More 'detailed discussion on - shutdown margin the economical analysis is presented in chapter 4. - power distribution - fuel design optimization claddin- g corrosion - hydrogen pick-up . EXTENDE3 D BORNUP IMPACNUCLEAE TH N TO R FUEL CYCLE dimensiona- l problems - fission gas release The economical aspects of extended burnup in different LWR reactors, - spacer grids R 100WE includin0d havan 0 e 44 gbee botR n WE hperforme d unde e IAEth rA - cladding collapse Water Reactor Extended Burn Up Study - WREBUS [5,6,7]. Included were also - fuel design - MOX, recycled U the main technical, licensing and environmental aspects. Therefore this - burnup impact paper deals only with main differences and requirements for more detailed - recycledU analyses. The possibilities of such analyses depend however on the - licensing problems decision e hackenth n o sd options, therefore both once-throug d closean h d cycles alternatives mus treatede tb resolutioe .Th n between main fuel cycle options wil e differenb l n differeni t t states e approacth , o suct h a h In distinction fro mR reactorWE PWR's r e problefo ,th s f claddino m g decision should be however more general and a matter for discussion. corrosion is not important, at least not for WER-440 (because of better properties of ZrlNb, lower temperatures and different water chemistry). e detaileTh d impacts coul dividee b d d into three parts- Most important will be the fission gas release limits for both WER-440 and - front-end, WER-1000 additionn .I WER-100r fo , reactivite 0th y coefficients, shutdown - design and operation, and margin and cladding collapse are to be properly analyzed. In more details back-end- . R thproblemeWE s were discusse [8,9]n i d . 3.3. BACKEND Closed cycle: designe th Simila r backende ,fo th als s r maie ra ofo th . n possible problems Spent Fuel Reprocess . to be solved in connection with extended burnup were discussed within the Discharge . W A R d an WREBUS Study (31: from Core Disposal Time, .t - spent fuel pool design - reactivity hea- t removal RA Waste Reprocess, - shielding Disposal and Pu, U - intermediate storage fuel facilit. y 2 a Fe - e credit3 a - - optimum technology, - reactivity - heat removal - shielding al [$/kgU Spen- J t fuel disposa time fuef th paie eo lt fe lda discharge claddin- g integrity from core - spent fuel transport a2 tf/kgU] - Ra waste disposal fee paid at the tine of fuel discharge from - reactivity core - heat removal a3 [$/kgU] - Reprocessing cost minus credit for Pu and U recycling {or plus - shielding penalty for their recycling) - final spent fuel disposal t [years] - Time between fuel discharge and final solution (either final - fuel reprocessing FA's disposal or RA waste final disposal) burnabl- e absorber impact i [55] - discount rate - RA waste storage fina- wastA lR e disposal e mosTh t importan clearls i t backene th y burnus dit cosd p an tdependence . Both cycles wil economicalle lb y equa: if l n cas i f o once-throug ed an h cycle scalinth e g facto s alsi f ro o consequence . al.U+i/lOO) a2.(l+i/100)*= t a3 (1) 4. BACKEND POLICY - MAIN PROBLEMS Solving this expression for t gives: As was shown in the WREBUS Study, the most important factor to consider for t = log f(a3/al)/(l-a2/al)]/log(l+i/100) (2) burnup extensio e backenth s i nd cost. However, backen s alse i pard th ot with highest both technical and economical uncertainties. As it was already discussed in Chapter 2, the economical comparison of both once-through and once th e timt th throug< e r Fo h cycl mors i e ee tim attractivth e r fo d ean closed cycle alternatives (because of different approach to payments: in > t the closed cycle is more attractive. the cas f finaeo l disposa paymente th l made advancesar n ei e cas th f eo n i , reprocessing - such payments are in the time of fuel arrival to the The possible range for a2/al is 0.1-0.9 (the final disposal of RA waste reprocessing plant and uncertainties) is very difficult. from reprocessing must be cheaper than final disposal of spent fuel). Let us assess the two alternatives: The possible range for a3/al is 1 - 10 (if the ratio is < 1, the fuel reprocessing will be always cheaper). Once-through cycle: s implicitlIi t y assumed, thar reprocessinfo t g either differens i e fe t set, or the difference will be made available to utility in the time t in the corresponding amount Kal-a2). (l+i/100) 1. Time, t show7 Fig exampl.n sa giv9 parametrie Figd d th eean an . 8 c evaluatiof o n the equation (2). This understanding is very important for both the utilities to make proper decision and for the state policy by proper setting the backend fee, for not to push the utility to solution which could be, from global point of view, less economical. k$/kgü S/kgU 1000 0 10 0 9 0 8 0 7 0 6 0 5 0 4 0 3 0 2 0 1 0 10 20 30 40 50 60 70 80 90 100 Time, years Time, years SF 1000 S RA 333 S SF 1000 S RA 333 S —— SF disposal - RA waste d. —— SF-RA —— SF disposal - RA waste ü —— SF-RA OJV Rez OJV Rot Fig . 7 Forwar. Backward an d d Discaunted d an ValueF S f o s RA Waste Disposal Cost a3/al Time, Years 100 90 80 70 60 50 40 ÎO 20 10 0 10 15 20 25 1234667 9 10 a3/al Discount rale a2/al 2 6 -0.1 ——0.2 0.4 ••••0.« ——0.« ——0.9 10 14 UJV Rei Fig . 8 Dependenc. f a3/al/(l-a2/alo e ) Fig. 9. Minimum Storage Time for on realtive backend costs Reprocessin Economicae B o t g l (a3/a a2/ald an l ) 43 5 COHCLUSIONS NEAR-TERM PLANS TOWARDS At present the average fuel burnup in Czechoslovak WER reactors is 30 INCREASED DISCHARGE BURNUPS MWd/kgU wit yea3 h r cycles Ther e incentivear e o implement s yea4 t r cycle IN FINNISH POWER REACTORS with burnu MWd/kg0 4 p- U Further burnup extensio f interesto s i n witht ,bu Review of background aspects and current activities the understandin e maith n f o constraig e leas experimentath FA d f o n l verification S. KELPPE Slightly different situation wil e witb l h WER-1000 These reactore ar s Technical Research Centr f Finlandeo , suppose staro t d t wit yea3 h r cycle d correspondinan , g burnu 0 MWd/kg4 p~ U and the perspective of extension to ~ 50 MWd/kgU with 4 year cycle, is Espoo, Finland already formulated Abstract REFERENCES Fuel cycles for the two reactor types used in Finland, 2 x WER-440, 2 x (1) IAEA International Conference on the Safety of Nuclear Power Strategy ABB Atom BWR describede ,ar e currenTh . t bur limitp ne u th d som san f eo for the Future Vienna, Austria, 2-6 September 1991 technical aspect nease th affectinr n i futur d limite e an egar th w sno reviewed. Examination researcd san h tha Finise tth h power companies 12 PazderaF ] JakesovL , â Czechoslovak Fuel Cycle Back-End Strategy increaso t carr t you e high bur datp nu a bas listede ear . Questionnary for Basic Data Collection Final (Version 1) UJV 9318 T, Rez, February, 1991, Proprietary LangM P Pazder F ,] [3 a Status e IAEA'Reporth n so t WREBUS Study IAEA Technical Committee Meetin n Fueo g l Performanc t Higa e h Burnur fo p Water Reactors Studsvik, Sweden, 5-8 June 1990, IWGFPT/36 1 Introduction [4] V V Yakovlev, F Pazdera, V D Onufriev, V N Proselkov, 2 Valvoda, Yu K Bibilashvili Incentives for Extended Burnup for WWER Reactors 1 Nuclea1. r Powe n Finlani r d Technical Committee Meeting on Fuel Performance at High Burnup for Water Reactors Studsvik, Sweden, 5-8 June 1990, IWGFPT/36 In 1990, about 35 % of the Finland's electricity production was covered by nuclear power. The generation is [5] IAEA Water Reactor Extended Burnup Study IAEA, Technical Series shared between two power companies Imatran voima Oy (IVO), Report, in print with its two 465 MWe modified WEE-440 PWR units {Loviisa 1, 1977; Loviis 1980), 2 a Teollisuuded ,an n Voim y (TVOO a ) (6] F Pazdera Preliminary Evaluation of Fuel Cycle Alternatives for CSFR BWRe MW operatinS 5 (TV 73 , 1980 1978, o II OI tw gO );TV UJV 9514 T,E, Rez, November, 1991, Proprietary Czechn (i , ) built by ABB Atom AB. The accumulated experience currently amount o som t s0 reacto 5 e r year totaln i s e operatioTh . n [7] F Pazdera, V D Simonov, V N Prosyolkov, V V Yakovlev, Z Valvoda historie l thesal ef o s units have been excellente Th . USSR/CSR Results for the IAEA WREBUS Study UJV 8871, Rez, 1989 average load factors for the units have amounted to some 90 % for the recent years. The annual collective radiation StrijovN P DubrovinP ] K ,[8 YakovlevV V , PazderF , a Computed an r dose n workero s s have been aroun e manS2 d th r eac fo f vo h Experimental WER Fuel Rod Modelling for Extended Burnup Technical two sites. Committee Meetin n Fueo g l Performanc t Higa e h Burnu r Watefo p r Reactors Studsvik, Sweden Jun8 ,5- e 1990, IWGFPT/36 The Finnish reactors are run in base load mode. Large seasonal variations are typical of the capacity demand. PazderaF ] [9 Bart,0 a Researc WWEn ho R FueBehavioud Ro l Operationad an r l Abundant hydro power and lowered demand cause spring and Experience ANS/ENS International Topical R FueMeetinLW l n o g early summer to be naturally favourable times for Performanc 90'e e th s Fue r Avignonfo l , France, April 21-24, 1991 refuelling. A principle application to the Finnish Government for buildin a gfift h nuclear uni f somo t n capacitei 100e MW 0 y e powes placeth wa ry b d companie y 1991Ma .n i s Offerr fo s seven concepts from three suppliers are being currently assessed. Statement from the Government, followed by the decisive Parliamentary resolution may be expected in 1992. 2 The fuel cycles 2.2.1 Front end (Loviisa) 2.1 TVO I and TVO II (twin ABB Atom BWRs) e utilit o WER-44Th tw O procure e IV y th 0 e fuer th sfo l units solely froe originath m l vendor e latteth , r being 2.1.1 Front end (TVO) the only manufacturer of WER fuel so far. The fuel and assembly characteristics are standard of the manufacturer; O procureTV varioue e uraniuth th s d san me stepth f o s two or three fixed enrichments are offered. Partly based front end fuel services (uranium, conversion, enrichment, e feedbacth n o k froe Finnisth m h experienc d IVO'an e s fabrication) separately with the fuel tailored to meet the suggestions, however, the manufacturer has introduced needs. TVO's polic bees ha yn towards diversified supply, several improvement e fueth l o t sove r longer terme Th . the fuel being now manufactured by ABB Atom AB in Sweden e fue% enrichmen wits th buli 6 l f 3. ho k d usen an ti d y Siemensb d an , Germany e planninTh . f fueo g l management three-year residence time mode. Some bundles, including is jointl e utilityyth carriey b e vendor y th ,t b ou d an s the contro followersd ro l e onlo years,se tw y ' life. the Technical Research Centre of Finland (VTT, independent national laboratory). Detailed non-destructive and destructive examinations on fresh fuel from each delivered batch have been carried out w bein planI almosn no O ru s gi tTV t entirely wite th h iy contracnb FinlanT VT y tb d from IVO. Thi s beesha o t n Siemens 9x9- e fue1th lfuel, typII , O ewhilTV n i e substitut e limiteth e d possibilitie o perforst m quality currently prevailin Atom'B l typesAB al s n i sgI , SVE. 64 A control e auditfueth lt a sfactory neee b dt . no Ther y ma e axial natural uranium blankets are being utilized for y mor an o routinelt e y continue this practice. improved axial power distribution. Burnable gadolinium absorbers are utilized in both units. Deviating from Recently, ther s beeha e n some interes y otheb t r companies one-year refuelling interval is not considered appropriate in lookin optionr fo g o star t sE fuetWE l manufacturiny b g in conditions that are expected to prevail in the near other companies. Siemens r instancefo , s recentl,ha y future. Low-leakage (in~out) loading strategies are listed a WER design fuel assembly with 2r-4 cladding. followed. In reloading, about one third or one fourth of the core is replaced in Loviisa and TVO reactors, 2.2.2 Operation (Loviisa) respectively. The average linear heat generation rate is 14.4 kW/m, permitted local maximum 32.5 kW/m. The fuel rods have been 2.1.2 Operation (TVO) prepressurized to 6 bars helium. Coolant pressure is 123 bars e claddinTh . g materia s recrystallizei l d Zrl%Nbe .Th The average linear heat generation rates are 14.1 kW/m for pellets feature a central hole of nominally 1.4 mm in TV I (9xO 9 fuel d 17.permitte; )an II 6 O kw/TV r dm fo loca l diameter e dominatinTh . g enrichmen. Fue* l6 3. s i t maximum is 41.5 kw/m. The currently utilized fuel rods are management plannin s mad i gy IVO b e . prepressurized. The cladding material is recrystallized Zircaloy-2. o burnablN e absorber e use vVER'n e momentar si d th t sa . The introduction of gadolinium for that purpose has been Fuel performance has been good. Over the about 25 cumula- studied. It proves possible to be able to do this by the ted reactor year sw failurejusfe a t s have occured. None end of 1990's. of the failures can be related to a burnup-dependent mechanism. Fuel reliability has been good. The rod failure rate is estimate t 4-10"a d 5 ovee accumulateth r 5 reacto2 d r years. 2.1.3 Back end (TVO) 2.2.3 Bacd (Loviisaen k ) After some cooling time the spent TVO fuel is stored in the interim storage planth t a esite e storagTh . s i e According to the initial fuel delivery contract, all the planned to be able to accommodate the spent fuel from the spent fuel assemblies from the two Loviisa units have been whole life time of the two units. The facility is planned and will be transported back to Soviet Union, the for storag e0 years 4 tim e desigf o eth , n life bein0 6 g ownershi d responsibilitan p y being rendere e borderth t a d. e finayearsth lr Fo disposal. o decisio,n s beeha n n made. e transfeTh r takes place after about five years' cooling The most probable solution is direct disposal in the e spenith n t fue e sitel th poo t .a l Typica t thelo l n Finnish granite bedrock. Several site candidate e beinar s g consists of the spent fuel corresponding one reaload from Öl studied. each of the units. 3 Genera2. l economy advanced fuel types. Introduction of those will bring the O) desirable burnup mark well over 40 MWd/kgU. With great uncertainties in the future nuclear policy and reactor types and in future pricing of the more advanced Possible introduction of a completely new reactor type in fuel types, only comparativ d highlan e y speculative cost Finland would start another deep-going procesf o s survey feasiblee ar s . Only after severae boundarth f o l y optimization and licensing. Issues of licensing play conditions having become more fixed can any optimation of naturall n importana y t role alread n advanci y e whee th n the future fuel cycles be accomplished on any realistic or offered candidate e comparedar s . absolute level. In general, the once-through fuel cycle is less sensitiv o dischargt e e burnup than that including high-cost reprocessing. Nevertheless, going higher in burnups is judged slightly but clearly favourable in Finland under condition e nowse .n sca thae on t 3.2 Technical fuel factors during operation 3.2.1 TVO fuels 3 Licensing issues One of the factors that contributed to the current burnup 3.1 Current burnup limits limits was the high fission gas release fractions found in B AtoAB me fueth som f lo e rods. There were also The national licensing authority, The Finnish Centre for indications that uncertaintie d locaro n li s power Radiation and Nuclear Safety (STUK) places upper limits estimates were larger than anticipated. BWR fuels in for the burnup individually on each fuel type utilized. genera le pronb see o o fairlt t em y large scatten i r fission gas release fractions. This may result from local e Loviis R th fue WE f o e lFoa th rplan maximue th t d ro m power variation o controt e du sl blade movementsr fo , averag 8 MWd/kgU4 t a e t burnu, se wit s n additionai pa h l instance, whic e difficular h o handlt n calculationsi e , requirement that no more than 12 assemblies may contain a and possibly from thermal feedback phenomena inside the rod averaging ove 4 MWd/kgU4 r . rod due to relatively low pre-pressure levels typical of R fuelBW s [1J. For the TVO fuels, there applies an assembly average burnup 6 MWd/kgUlimi3 f o t , additional restriction being n additionI e effecth , f channeo t l bowin n locao g l power that no more than 10 assemblies may exceed average burnup (via causing deviation e wate th p betwee ga n ri se fue th nl 4 MWd/kgUo3 f . channele resultinth d an s g combined effect f altereo s d thermal and fast neutron fluxes) have received attention During the 1980's, the allowed discharge burnups increased apossiblsa e high burnupe s givefactorth ha p u nO .TV 0 HWd/kgU/year1. o t 5 b0. y e fue.Th l performanc s quitwa e e earlier practice of reusing the channels. good and no burnup related failures were found. STUK did, however, froze the burnup limits at the level reached in TVO expects that the accumulating experience will show the 1989, maintaining tha o comprehensivn t e assessmene th f to scatte n releasi r e e e suppresserateb th o t o st e du d safety significance of increased burnup, particularly in adopted in-out loading strategies that curb large power relation to transient and accident conditions, exists. variation o movinst latee du n glife i roved o fuean ,t r l designs with lowe d linearo r d r an power 9 9x so t (fro 8 8x m In the near future, the Finnish power companies will possibl o 10x10).Normat y l operation experiencd an e evidently continue pursuing slightly higher burnups that extended irradiation programs suggest that cladding are readily achieved technically wite currenth h t reactors outside oxidation is probably not a major concern for the and fuel design and management concepts. This might mean burnups foreseen. little e currenchangth n d maximu i e8 ro MWd/kg t4 f r o m fo U the Loviisa WER fuel, provided the three-cycle mode will Going to high burnups PCI needs to be looked at with any be followed, and maximum assembly burnups of some fuel type. Still, for example, it has been seen that 40 MWd/kgU for the current TVO fuels. essentiall o ridginn y e claddings formeth ha gB n AB o d f o s 8x8 rods with burnups up to 48 MWd/kgU. In destructive Somewhat later on, there may be incentives to change over examinations, only slight bonding betwee e pelleth nd an t to four-cycle mode in WER reactors. This would entail a cladding was occasionally found. Fuel manufacturers claim profound re-evaluation of the fuel cycle economics, which improved PCI resistance in the future fuel types t beeno n s performeha O yet O wilIV TV .y lb d consider more considere y TVOb d . 3.2.2 Loviis R fueWE a l Ther e severaar e l safety factors whic he takeb hav o t en whict bu car hf o ehav e straightforward tehnical solutions, Ovee lif th re Loviise th tim f o ea R plantsfueWE e l th , the issues being rather of economical character. An has gone through significant development. One of these has exampl e enrichmen th f suco es i h e tfue th limi lr fo t beee pelleth n t manufacturing that, insteae formeth f ro d pools. Going somewhat e highecurrenth f o r t initial extrusion method w accomplisheno s i , y morb d e standard enrichments will need modification e storagth n i se racks pellet compressing technique allowing better command over and relicensing. The utilities also acknowledge that the desired properties. Improvements that potentially have higher discharge burnups quite rapidly increase the a direct influence on the high burnup performance include neutron radiatio n speni n t fuel, which e needtakeb o t sn allowing mor e e assembld axiaspacth ro n li r e fo ygrowth , into account in transport and storage. The interim and going from 1 bar helium filling to 6 bars. storages are licensed for burnups high enough to accommodat e foreseeth e n spent fuel batches. Also for WER fuel the licensing has been cautious. One of the worrie s beee fissioha sth n s releasega n . There ar e some early results from irradiations in Soviet Union that 4 Actions taken to qualify higher fuel burnups show unacceptably high release fractions. The experience from Loviisa plant suggests gas release fractions of less s basea n destructiv, thao % d 1 n e PIE r leso , s tha% 2 n e Finnisth Bot f o hh power companies carr t extensivou y e from poolside nondestructive determination (y-scan, Kr85). programme o collect s e tperformanc th dat n o a f theieo r Linear power Loviisn i s a WER-44 comparativele 0ar y low. fuel types. The Finnish licensing practice does not favour Pellet design wit a centrah l hole further contributeo st large-scale prototypic irradiation Finnisn i s h power w fuelo l maximum temperatures. reactors. To substitute this, the utilities go in for a numbe f actiono r s som f whico e e e listeth ar h n i d e higth h o t corrosio e Du n resistanc e claddinth f o e g following. material (Zrl%Nb), comparatively low cladding temperatures and favourable water chemistry, the oxidation on the used Loviisa claddings is almost negligible. Oxidation is 1 Poolsid4. e examinations expected to be of no concern even at considerably higher burnups. Both of the Finnish utilities have a versatile poolside fuel inspection facility which have been used very Destructiv s showha E n PI eclose d fuel-to-cla witp dga h activel r extensivfo y e routine th examinatio f o l al n no occasional strong bonding. It is felt that even fairly fuel types n mori s wela , es a ldetaile d surveys like those deeply depleted rods may not be safe from PCI-inducing mentione e sectionth n i d s below [3] e facilitie.Th s ramps especiall e neighbourhooth n i y f fresheo d r feature a number of standard equipments. In addition, more assemblies after refuelling. sophisticated examinations have been occasionally ordered from outside contractors. Examinations performed poolside The fuel failures have been given great attention by IVO in Finland include: [21. Leaking assemblies have normally been examined in the visual inspection of the assemblies and rod poolside stand. In a few of the cases, PCI failure is assembly distorsions suggeste r cannoo d t leasa t e overruledb t n somI . e other d lengtro h - measurements cases, there is strong evidence that the faults seen are clad profilometry secondary failures. Recently, distorsion e uppeth e ti f ro s oxide thickness measurement, fuel-to-clad gap plate have been detecte n severai d l bundle n visuai s l measurement with "squeezing" method a correlatioexaminations e b t no y n ma wit .r o Therh y ma e Y-scanning including fission gas release fuel failures. These deformations see o result m t from determinatio e Kr8th 5 y b activitn y method. jamming of the upper end of the axially growing rods to the upper tie plate. The design needs corrective modification. 4.2 Extended irradiations of single assemblies and detailed PIE examination f selecteo s dt cell ho rod n si s 3.3 Licensing issues for front and back ends (IVO, TVO) s carrien extedea ha t O ou TV dd irradiatio f severao n B AB l Atom 8x8 fuel assemblies. The irradiation lasted five and Withi e non-revolutionarth n y e foreseerangth f o e n burnup six cycles and the rod maximum burnup of 48.5 MWd/kgU increases e Finnisth , h utilitie e littlse s e technical (max. pellet 55.5 MWd/kgU s reached)wa . Nine rods froo tw m problems regarding front and back ends of the fuel cycle. of these bundles were included, after poolside examina- -pk tion, in the international Battelle High Burnup Effects co-ordination of the national laboratory VTT. Examples of 03 Progra detailer fo m E examinationPI d e Windscalth t a s e on-going or recently completed efforts are the OECD Halden Laboratories, England [4]. Valuable referenc s alswa e o Reactor Project, the many Studksvik fuel projects, the gained to calibrate the coming poolside examinations. For Battelle High Burnup Effects e ThirPrograth d an Rism a example, an excellent match was arrived at between the Fissio s ProjectGa n pasr w yearss obviouslFo .fe ha t t i , y fissio s releasga n e fractions measure y Kr8b d 5 method an d become more difficult to launch such programmes. This those from puncturing. Some of the conclusions are reflects the groving demand of more focused fuel-type referre n sectioi o t d n 3.2.1 above o clearlN . y restricting specific experiments. high burnup phenomena were found. MorE examinationPI e s (EPMA, burnup determinations) are being carried out at Windscale under commiss.ion from TVO. e rol f Th fueo e 5 l performance codes 3 Poolsid4. hotceld an e l examination Loviisf o s R fueWE a l Computerized fuel behaviour models are essential in understanding the many phenomena and interactions that in 1985-91 16 assemblies and five absorber assemblies have gover nucleae th n r fuel behaviour. Actually mose th ,t been examine e poolsidth t a d e inspection stand. During e experimentaefficienth f o e us t l result o condenst s i s e annual refuellings, the lengths of 47 irradiated a modee for th f theo m l o t mparamete a submode r o ra n i l assemblie d theian s o burnupr t 9 MWd/kgU rod4 p (u f so s ) code. Successful modelling is a sign of advanced have been measured [3], Four rods from three special high understanding. burnup assemblies were examined destructivel n Studsvii y k laboratories. More recently two special three- and The high burnups of the nuclear fuel place great four-cycle assemblies have been unloaded. These assemblies challenges to fuel modellers. Even currently, one can spot and some of their rods have been precharacterized in several weak r pointdescriptionou n i s e integrath f o s l detail. The poolside examinations of these assemblies and behaviour. rods will be performed in 1992. Rod burnups range up to 48 MWd/kgU. e phenomenth Fissio f o s release aga n on tha s i et remains difficult to predict. Occasionally, even the fuel vendors 4.4 Test reactor irradiations of WER fuel have had large scatter in simulating the behaviour of thei n fuelsow re probleTh . m s releaslie t onlga sno n i y e Under co-operation between the Kurtchatov Institute of model s sucha s e actua.Th l temperature d temperaturan s e Mosco d IVO an wn experimenta a , l progra s carriei m t ou d distribution f higo s h burnu e knowb t pn no fuey lma that produces highly qualified behaviour data from closely accurately enough. Further vers i yt i ,importan e us o t simulated WER-44 WER-100d 0an 0 conditionw sno (5Jo t p .U methods in power calculations that are sufficiently d bundlem activthre1 ro f 8 1 e(o es length) instrumented sophisticated. with fuel thermocouples and elongation sensors have been irradiated in the MR test reactor in Moscow. The fourth More mechanical data shoul e acquireb d d before detailed one, featuring alsd internaro o l pressure sensor s goini s g r clamodellino dI PC creep-ou f o g t (wit e possiblth h e into the reactor; fifth one is under construction. Among lift-off effect s actualizedi ) . the rods in the bundles, design parameters like fuel density, fuel-to-cla e variedar d file s an Th ga . l p ga d Still, ther s stroni e g nee f physicallo d y based fuel irradiation e followe t celar s ho l y b examinationsd . performance codes, thoug t looki h s like that certain calibration against each fuel typ n questioi e d an n In the second phase of the program, a facility will be detailed fuel-type specific materials data will be needed. installed and applied that allow reinstrumentation of high burnup rods with fresh thermocouples. This phase is In Finland, the larger fuel performance codes are run by starting in 1992. the national laboratory e frameworVTTth .n a I f o k national research programme, new codes are being acquired Plans also exisa thir f o td phase where ramd cyclinan p g and developed to extend the analytical capabilities into behaviour of WER fuel would be studied. appearin w rangene g f fueo s l desig d operatinan n g data. 5 Internationa4. l research programmes 6 Conclusions e FinnisTh h organizations have participatee th n mosi df o t internationally sponsored fuel research programmes. In Finlan e performancth d e fouth r f o ereacto r unitd an s Usually the participation has been actualized under their fuel s beeha s n very good. However, extremes a s regards reactor power densities and linear powers or discharge burnups have been avoided e licensin.Th g policy COMPATIBILIT EXTENDEF YO D BURNUP adopted by the Finnish authority STUK has been cautious. WITH FRENCH FUEL CYCLE INSTALLATIONS Extendin e dischargth g e burnup s judgei s d clearlyt ye , P. DEMOULINS maybe slightly, favourabl n economicai e d environmentaan l l terms unde e advocateth r d once-through fuel cycle concept. Service des combustibles, Electricité de France, e FinnisTh h utilities pursue slightly increased discharge Paris e nearesburnupth r fo ts futur r theifo e r fueln i type w no s use. P. SAVEROT e utilitieTh d carr an st remarkabl ou y e fuel examination Cogéma campaign f theio s r own. Introductio w fuene r o lf o n reactor types will bring incentives to further extend the France burnup limits. Abstract International co-operation is important for Finland in experimental fuel research. Based on the fact that about 75 % of national electricity suppliedis nuclearby power plants, economy Computer modellin hige th h f burnuo g p phenomena should of nuclear energy is a very important concern in France. keep up with the recent experimental findings. The effect At this level penetrationof takingand into accountthe of burnup on transient and accident behaviour and related experience gained through the 15 past years during t/hich modelling e forgotteshoulb t no d n either. about 50 plants have been constructed and operated, the only effective way for reducing the cost of nuclear energy appears to be burnup extension. This program is REFERENCES currently progressin involvesactorsand the the of all fuel cycle, from enrichment phase to reprocessing Persson, S. et al., Evaluation of some high burnup phenomena activities. The burnup extension program carried out in in boiling water reactors. ANS/ENS Topical Meeting on LWR Fuel France reviewed.is Performance, Avignon (France) April 21-24, 1991. Moisio, J. et al.. Experience of operation with leaking fuel at d examinatioLoviisan S NF a f leakino n g assemblies. ANS/ENS Topical MeetinR FueLW ln o gPerformance , Avignon (France) I - PREAMBLE April 21-24, 1991. Lindroth , TeräsvirtaH. , , PosR. , t Irradiation Examinationf o s countriel Inal s concerne nucleay db r energy, burnup extension High Burnup Fuel and Leaking Fuel of Loviisa NPS. IAEA TC is one of the main concerns of the plant owner. In fact, such Meeting on Poolside Inspection, Repair and Reconstitution of extension has a significant impact on the cost of electricity R FueLW l Elements, Lyon 21-24 October, 1991. produced. Cunningham, M. et al., Qualification of Fission Gas Release For reference, figure 1 attached to the present paper shows the Data from Task 3 Rods. Battelle High Burnup Effects Program, variation of fuel cycle cost in relation with burnup value. This HBEP-60 (3P26), Battelle Pacific Northwest Laboratories, figur s establishewa ebasie abouF th f EO so n o year3 ty db o sag January 1990. the economic conditions prevailing at that time, and may suffer some minor modifications linked with variations affectine th g Jakovlev, V., Johansson, J. et al., Research of WER-440 type constituting elementary costs. fuel rods in MR reactor. IAEA-SM-288/64, September 15 to 10, 1986 Stockholm, Sweden. However, the general shape of this figure is still basically correc d showtan s givea that r nfo , refueling policy, fuel cycle cost reache sa minimu m valu burnur fo e p rangino gt fro0 5 m -£>• 60 GWD/T. This can be achieved provided that fuel enrichment is (O . raise% t leasa 5 o 4. t d Öl In this respect F havEO ,e started since some yearo t s o progressively increas e fueth el enrichmen theif o t r domestic plant ; this s polic s illustratei e y th followin y b d g milestones : c/kWh Step Period Enrichment% B.U * GWD/T 1977 3.25 ( 900 MWe) 33 1 1984 3.10 (1300 MWe) 33 1987 MWe0 90 )3.7( 0 42 2 1994 3.60 (1300 MWe) 42 3 > 2000 5.00 60 (* Typical average unloading B.U. value) -wortIs ti h notin followinge gth : s - At the first step, EOF policy consisted in loading fuel elements enriche 3.2o t Thi. % 5p du s enable operato da t n o e 1/3 core yearly refueling basis. - More recently, enrichment has been raised to 3,70 % on 900 MWe plants, enablin burnue increasee GWD/2 gb th 4 d o o pt Tan t p du to operate on a 1/4 core yearly refueling basis. F o targeraist EO o t s - i es ta fueo s l enrichmen% 5 o t p u t reach a 60 GWD/T burnup. This challenge needs a lot of problems to be overcome. Experience gained during the 2nd step will be helpfull for that task ; it is the experience gained by EOF as operator and COGEMA as reprocessing firm in the framework of this first enrichment 20 30 40 50 60 70 80 GWD/T extension we are going to discuss here after. It covers the following technical aspects : F\G. . Fue1 l cycle cost versus burnup. enrichmen- fued an tl manufacturing - transportation of fresh fuel elements - storage of fresh fuel elements cor- e reactivitd loadinan p gma y control - storag spenf eo t fuel elements - evacuatio transportatiod nan spenf no t fuel elements - reprocessing fully covered by the French industry. 2 - ENRICHMENT AND FUEL MANUFACTURING Maximum Brand enrichment with Absorbing Fuel type % material nowo t ,p U uranium enrichmen d fuetan l elements manufacturing for national needs have been mostly carried out by French firms. Yearly national needs, including first core loading and FRAMATOME RCC3 3.9 Cu 900 refueling, are summarized in the following table (year 1991) : FBFC RCC 3.8 Cu 900 - Natural uranium needs 8 000 T/year - Enrichment capacity SWU/yeaM 6 r FBFC RCC4 4.5 B 1300 Manufacturin- g 2560 fuel assemblies/year With respec enrichmento t , whic carries y i hb EURODI t dou n i F capacite Th thesf yo e container elements2 s si . the Georges Besse Plant at Pierrelatte, enrichment rate up to 5 % is allowed by French Authorities regulations for public These containers fulfill EDF needs for several years ; further needs. This limit is consistent with EDF forecast needs up to developments will have to be carried out for facing the problems programe th f o .) % 5 e ste( th 3 p linked withighee hth r enrichment forecast. It is worth noting that small quantities of uranium enriched at hav% 5 e 4. already been produce thiy b d s diffusionF planED r tfo needs. STORAG- 4 FRESF EO H FUEL ELEMENTS s regardA s manufacturin f fueo g l elements, this activits i y As regards the storage on site of fresh fuel elements, the only carried out in France in accordance with regulations issued by problem which was identified was that of criticality. the French Government. These regulations specif maximue th y m enrichment allowed for the uranium to be used by the Radiation shieldin raist problemno y ed an gdi . manufacturer. This enrichment is fixed at 5 % for the ROMANS factory (decree Firs allf o tt musi , kepe mintn b i td o policthaF t ED ts i y store fresh fuel element plane th t n si fue l building poolo ,s dated 2 nd march 1978) and the PIERRELATTE factory (decree dated that the problem is to determine if there is any risk of 7 th sept. 1982). criticality (keff < 0.95) for underwater storage. In this This limit match wit F r goastehED fo l3 p as defined here respect, it has been determined that this condition is fulfilled above. by fresh fuel element whe% 5 ns 4. enrichestore o t n i p du d stainless steel racks in non borated water if the cell pitch exceeds 380 mm. For 900 MWe plants, and for the first 1300 MWe series (P4) , this thao s proble o tn , mm 0 m41 pitc wil s i lh occu thin ri s fieln i d TRANSPORTATIO- 3 FRESF NO H FUEL ELEMENTS neae th r future. For the next series (P'4 and N4), the pitch is reduced to 280 mm Fresh fuel elements transportatio regulates i Frence n th hy b d (compacted spent fuel storage). These compacted racks meee th t Authoritie conformancn i s e wite domestihth d internationacan l criticalit % enriche 5 4. y o dt requiremen fuep y u b l r fo t regulations applicable. incorporating neutron absorbant ; this was taken into account at the design e seriestagth f eo s concerne thao s d t extending fuel Fresh fuel containers shall be approved e Ministrth by y enrichment will not require any re-racking operation until the responsible on a case by case basis. step 3 of the program. maie th n f problemo e On s raise enrichmeny db t extensio than i n t field is that of criticallty ; for that purpose, fresh fuel containers shall be equipped with absorbing materials (copper, COR- 5 E REACTIVITD LOADINAN P HA G Y CONTROL borated steel). 900 MWe plants were initialy loaded with 3.25 % enriched fuel e timAth t e being type3 , f authorizeso d container e user sar fo d and 3 coroperate1/ ea yearln o d y refueling basis. Core loading EDF need specifies sa e followinth n i d g tabl: e map was according to figure 2 ("Out-In-In scheme"). rÖol HGFEDCBA HGFEDCBA 8 Batch? Batch 4 Batch8 Batch 5 Batch9 Batch 6 Fresh Fresh FIG. 2. 900 MWe core loading map: 1/3 core 3.25% V5. FIG. 3. 900 MWe hybrid core loading map: 1/4 core 3.7% U5. Whe nyeae th renrichmenn i 198% thre7 n 7o raises 3. twa e o t d experience gaine thin i ds field demonstrates that gadolinium plante MW 0 s 90 (Blayai Graveline, sl Chinod an sl n Bl) same ,th e burnable poison is a reliable technology for future enrichment fuel management led on one hand to change the core load map to a extension. "hybrid" one in order to reduce the core hot point factor in some situations and to limit the neutron fluence on the reactor Otherwise, initial boron concentration and control rod e otheth re n burnabl o us vesselhand o d t ,an e, poison worth shall be checked every time enrichment is modified or fuel (gadolinium) in order to control the reactivity. managemen changeds i t s .A regard s boron concentrations i t i , mainly connected with cycle duration ; in this respect 1/3 core Actually policF operato e reactort EO th , s i l y eal s loaded with 3.25 % and 1/4 Core - 3.7 % schemes lead approximately to the 3. % 7enriche 4 cor1/ de a fuebasi n o l unde; s r this condition, same boron concentration requirement othee th rn .O s hand ha t i , burnable poison is no longer mandatory. In fact, it has been been determined that higher enrichment would possibly require determined that reactivity control only requires in these enriched boron control rodr reactivitsfo f o d y en contro e th t la circumstances absorbing Pyrex rodfirst a s t core loadings A . the fuel cycle. regard core sth e loading map "hybridn a , " schem s beeeha n set-up (see figur. ) 3 e Investigations carried out in the framework of the 2nd step project have demonstrated that fuel enrichment extension up to 3.7 % does not induce significant changes in the core management scheme. Higher enrichmen d fuetan l management 3 likcor1/ e would require burnable poison to control the reactivity ; 6 - STORAGE OF SPENT FUEL ELEMENTS On the basis of a 2 years decay period, as discussed here after takind ,an g into accoun tfuea l management operating on a 1/3 or 1/4 core yearly refueling basis, the Spent fuel storag n siteo e raise s4 kind questionf so s which operational lag storage capacity should be about one core mus answeree tb d when extendin burnue gth p valu: e for each plant. - radiation dose in the plant fuel building This conditio F r plant EO s fo fulfillei n l : s al y db criticalit- y reference, CPO desactivation pool capacity enables to - residual heat removal store 2 cores ; this capacity has been extended to 7/3 - operational lag storage. core for CP1 and P4 series and to 3 1/4 core for P'4 and N4 series. Radiatio- 1 6. nservice dosth n eo e e fueflooth lf ro building kepe b tio st belo e "greewth n area" upper limit valu5 (2 e /j Sv/h) under any operating condition. EVACUATIO- 7 TRANSPORTATIOD NAN SPENF NO T FUEL ELEMENTS In fact, it was computed that the water depth within all the pools concerned by spent fuel handling and/or storage s largwa e enoug accomodatinr fo h g high activity fuel Evacuation of spent fuel assemblies froa the plant desactivation elements. For reference, the dose criteria quoted above poosubmittes i l conditiono dt s activite linkee th th o f dt o y requires a minimum water thickness of 255 mm shall be assemblies, their residual hea theid tan r reactivity. maintained betwee watee nth r a e surfachea th f o d ean fuel element enriched at 4.5 % and irradiated at 45 GWD/T. Burnup extension, and Correlative high enrichement have obvious Actually, the water thickness in EOF plants is in no case effects on these parameters. smaller than 3 m. However, special measures shall be taken as regards spent As regards the residual heat, though the difference between a fuel element sviscinite storewatertighe th th n i df yo t 3.2 enriche% 5 d fuel assembly% irradiate5 4. GWD/a 3 3 d t Tan da doors of the spent fuel pool : operators have been enriched fuel assembly irradiated at 45 GWD/T is quite small instructed not to handle and not to store spent fuel during the first days following reactor shutdown, it becomes withi shora n t distance frodoorse th m . Forbidden storage significant afte monthr6 s (see figur appended)e4 . cells concerned may be loaded with fresh fuel assemblies. 6.2 - Taking into account the measures taken against the risk of fresh fuel criticality (see sectio spen, n4) t fuel storage does not raise any problem in this field. RESIDUAL HEAT (W/T) COOLING ) (% P A 6.3 - Figure 4 appended provides a comparison as regards TIME 3,25 % 4,5 % residual heat between two significant cases (3.25 % - 33 MWD/0 300 3 T 45 000 MWD/T GWD/T and 4.5 % - 45 GWD/T). The power developped within 4 days following reactor shutdown is only given for consideree b o t informatiot no d s whei d n an dimensioninn g 1 h 5,68 ES 5,82 ES 2,4 the residual heat removal system (PTR) of the plant. 1 d 2,41 E5 2,55 E 0 4,16 4 d 1,45 E 9 1,56 E5 4,7 e worsth Int situation (i.e .4 day s after reactor 15 d 8,10 E4 8,79 E4 8,5 unloading), ther onls i esmala y l- differenc % 7 4. - e 1 m 5,94 E4 6,54 E 5 10,2 between the residual powers relieved by the both types of 6 m 2,11 E4 2,51 E4 25,8 assemblies considered. This variation can be accomodated 1 Y 1,20 E4 1,51 E4 25,8 R systePT e m th withou y b y difficultyan t . 1,5 y 8,63 E3 1,07 E4 28 2 y 6,23 E3 8,30 E3 33,2 3 y 3,83 E 7 5,38 E3 39 e requirementth o t e Du - s presente4 6. hige th h y activitb d d an y 5 y 2,11 E3 3,14 E3 48, 8 e heath t loa f speno d t fuel discharged froe reactoth m r cn (see sectio hereafter)n7 operationan a , g storagla l s i e CJ required. FIG. 4. Spent fuel residual heat. othee th r n handO e Oinitiai,th l enrichmentakee b no t ints i to - Installations were designed for a full cask thermal ** accoun regards ta rise scriticalitf th ko y during transportation . Kw powe5 8 f ro reprocessine th t ana d g plant (reprocessin damagea g d assembly, weak irradiatio e uppe th e assemblies)th rf o n parf o t . wells A e initia- ,th ln a desig base f mads o th e nwa n eo Similarly minimu,a m burnup valu requires assembliee i e th r fo d s initial enrichment of 3.5 % and a burnup value of to be evacuated so as to reduce their reactivity. 33000 MWD/T. Consequently, spent fuel evacuatio subjectes i n n Francdi o et time Atth e being, higher initial enrichment together with two kind requirementf so : s higher burnup are accepted, provided that : - it is checked that the spent fuel assemblies have A - Spent fuel cask requirements actually been irradiated during one cycle, at least, - Radiation dose at the surface of the spent fuel cask - the spent fuel assemblies have stayed in the shal Sv/e lest b a f ls h t 0 a thaSv/hm 10 d n2 an , desactivation pool for a minimum period of time, as set distanc. m 1 f eo forth here after : Therma- l powe assemblief ro s enclosed shal lowee lb r than a specified value. BO GWD/T 35 40 43 45 Initia- l enrichment shal lowee b l r tha a specifien d value. 900 (days) 270 310 420 550 Spent fuel casks useneed F Francn e EO i dliste ar sr efo d 1300 (days) 320 370 470 580 hereafter, together with their specific characteristics. These requirements have two major consequences : NPP Max. Power Max. BO Max. Enrich. Type Brand (kw) (GWD/T) (*) s regardA - decae th sy time required ,a limitatio f no burnup extension will rapidly be created by the plant desactivation pool capacity. Externa g storagla l e 900 TN 12/1 120 42 3.75 facilitie necessare b y sma than yi t case. 900 TN 12/2 93 42 3.75 - With respec burnue th o pt checking requirement, French 45* firms have already developed a system enabling to monitor spent fuel assemblies in reactor pool or in 900 LK 100 B 85 42 3.75 reprocessing plants with N.D.A methods prototype .Th e device called PYTHON enables by passive and active 1300 TN 13/2 109 42 3.75 methods (neutron and gamma counting) to measure the 3 45* main physical parameter fuea lf so assembl burnup: y , cooling time and effective multiplying factor. A first industrial system is today in operation in France. * Over-shielded type This table shows that no spent fuel cask is at the time being qualifie transportatior fo d ovef no r 3.7 5enriche% d 8 - REPROCESSING fuel. This activity is carried out by COGEM thein i A r factort a y Reprocessin- B g plant requirements La Hague. The following data are those applicable to La Hague For the purpose of this report, high burn-up fuel is defined by reprocessing plant. comparison to the characteristics of the nominal UP3 fuel (i.e. TABLE 1 observe% mor 0 4 e thers e a fissiodar e n products % mor 5 ,2 e plutonium, an increase of 100 % in the alpha content as well as HBU FUELS - COMPARED CHARACTERISTICS an increase by a factor of 3.5 of the neutron emission (due to ESTIMATED VALUES Cm 244) . A summary listing of the impact of these new fuel characteristics on the reprocessing plant is as follows : Data refers to 1-metric ton of initial uranium (MTU) - increase of the U235 initial enrichment : criticality impact from the storage pool NPH to the Rl buffer tank unit, DATA NOMINA3 UP L HBU XCOEF. OBSERVATIONS Burn-up 33,000 HWd/HTU 47,000 HWd/HTU 1.4 TABLE 2 Cooling time 3 years 5 years 1.66 U235 initial 3.5 % 3.7 to 4.5 % COMPARED CHARACTERISTICS OF NOMINAL UP3 FUEL AND HBU FUEL Fission products : 35 kg 49.g k 5 1.4 Including solubles 26 DATA NOMINAL UP3 FUEL HIGH BURN-UP FUEL insolubles 3 gazeous 6 Burp n-u 33,000 HWd/MTU 45,000 MWd/MTU FP activity 7.4 x 105 Ci 6.5 x 105 Ci U235 3.5 % 3.7 % Cooling Time 3 years 4 years Activation products . hull % 0 +2 ACTINIOES . end fittings ]lZ to 14,000 Ci % 0 +2 1.2 . mass U 955.g 4k 940.8 (0.985) Actinides : . mass Pu 9.74 kg 11.37 (1.17) . U + Pu 963.14 kg 952.17 (0.987) . Uranium 232 0.090 ppm/U235 0.150 to 0.220 . beta, gamma 1.22 105 Ci 1.5 105 (1.23) ppm U235 . alpha 7.057 103 Ci 1.4(2.1« 10 7 ) . Uranium 235 0.9 % residual U 0.8 « residual U depends on initial hea. t release 237.3 W 500.1 (2.1) enrichment . neutron emission Pu 3.058 108 n/s 1.068 109 (3.5) . Uranium 236 0.5à 4 %0. resU . 0.6 f. residual U idem - penalty for U alph. » activity 3868 Ci 6608 (1.7) hea. t releasu P e 126.3 W 216.7 (1.7) Total Plutonium 8.9 kg 11 kg 1.25 neutro. n emissiou P n s n/ 7.19 6 10 4 1.162 10' (1.6) including : . % Pu 238 1.8 % 2.8 8(1.6% ) 8 23 u P . 2 % total Pu « 5 tim3. . 1.75 . 7. Pu 240 22.7 % 24. (1.08% 6 ) totau P l 1 24 u P % . 12.2 % 12.8 % (1.05) p N . 483 g 700 g 1.50 . Am 257 g 600 g 2.75 . alpha activity/kg Pu 397.2 Ci/kg Pu 581.1 (1.46) . Cm 29g 80 g 2.50 estimates . thermal power/kg Pu 12.97 W/kg Pu 19.06 (1.47) 4 24 m C . 24.03 g 82.46 (3.43) Total activity 6,73i 6C 13,000 Ci 2.00 r fineZ s 1kg 1 kg 1.00 ACTIVATION PRODUCTS . total mass 316 kg 316 (1) . activity 6193 Ci 5735 (0.93) 33,000 MWd/MTHM U23% 5 5 3. ,initia l enrichment year3 , s cooling . Heat release 37.22 W 38.23 (1.03) before reprocessing) as follows : . Co 60 2261 Ci 2366 (1.05) - a typical value of 47,000 MWd/MTHM, - an U235 initial enrichment between 3.7 % and 4.5 %, FISSION PRODUCTS - 5 years of cooling time before reprocessing. . mass FP 33.8g 2k 46.07 (1.36) The main characteristic f stheseo referenco tw s e fuele ar s . activity 761500 Ci 747400 (0.98) . heat release 2792 W 2692 (0.96) compared and estimated in Table 1 while Table 2 compares the . Ru/Rh 106 i C 4 10 4 6. 4.61 4 (0.7210 4 ) cn values of the nominal UP3 fuel with those of a high burn-up fuel 5 8 r K . i C 3 10 9 7. 9.33 (1.1910 7 ) 01 after onl year4 y coolinf o s g time. Major differencee b n ca s increas- totae Oith lfissioe f eo masth f so n product impac: s n to TSHEARIN- l Dissolution-clarificatio: G n concentratioP F e th vitrificatiod nan 5 0 n units, - The main problem is located at the clarification unit which e - increastotath lf o mas ef plutoniuo s e impac: th m n o t has to process 30 % more fines, thus needing more frequent throughput capacit 4 plutoniuR f o y m unid BSTlan t , criticality cycles. This limitation should constitute one of the impact due to the increase in Pu concentrations, bottleneck alonl planse al gth t process. alphu P - increase a th n activiti e compare: ye fissio th o nt d products activity, this leads to a decrease by a factor of 1.7 Hig- 2 hT activity solvent extractio wastd nan e concentration ratie th ofo beta activit yalph/ a activity wit possibla h e impact on the definition of the monitoring equipment, Dependin- limitine th n o g g value which wil e determinelb r fo d the Pu/U ratio, HBU fuel can or cannot be processed on line in - increas d neutroCuriun an i e4 24 mn emissioe impac: nth n o t this facility . a limitin This i s g point fro a licensinm g threshold atmospherie valueth r fo s c release estimateo t e sdu standpoint. e importancth Curiuf eo thein i m r determination. Impace th n to neutron emission of the fission product solutions, due to - Apart from this e otheon , r important incidence b n ca e Cm 244 ,e biologica th thu n , so R7 l , shieldingR2 , Rl f so evaluated V quantitieconcerninI U e th gsu P neede r fo d SPF., extraction, which could be also a limitation for the overall capacity. - increase of the Pu specific thermal power : impact on the temperature of Pu solutions in R4 and of PuC>2 in BSTl and on - FP concentration will have to process 40 % more fission the equilibrium temperatur FS4e th 7 f transporeo t containers, product y increasin(b s g concentrations and/or volumef o s solution handle)o st . - increase of the Pu neutron emission : impact on the R4, BSTl biological shielding. purificatiou P d an u T- 3 n Basically, two ways are possible to reprocess high burn-up fuels throug e reprocessinth h gn lino Throug) plan1 ee : th th - Although not directly a process limitation, the evolution of reprocessing of batches of segregated high burn-up fuel which the isotopi consideree cb compositioo t s ha : d U f no implies some new licensing issues, 2) Through dilution of the high burn-up fuel at the earliest possible stages of the . increas f irradiatioo e o U23t 2e ndu increas e (neef do process. complementary shielding), Dilutio higa f hno burn-u nominape th fue n i ll fuel mus done tb e . the penalty due to U236 (neutron absorber) significantly procese th earl n i sy i.e. just afte dissolutioe rth ne steth : p increased. o dissolutiotw n coull lineT f dso work simultaneously (one line for the nominal fuel, one line for the high burn-up fuel). purificatiou Withi- P e th n n line significan,a t increase th f eo Furthermore buffe,a r tank coul e installedb 2 dT between a l nT irradiation level is caused by the increase of Pu238, which o t receiv accound ean hige th ht burn-u po keet e fued pth lan requires more shielding. This irradiatio causa s i f fasteneo r high burn-up solutions in stand-by until they can be fed in the damagin solvente th f o g , reducin efficiencys it g . main flow without throughput capacity reduction of the plant. The successive UP3A facilities are listed below along with the T4 - PuC>2 conversion and packing main consequence n so lin froe e th mreprocessin batchef go f o s high burn-up fuel : e reprocesseb o t mor% u 5 P e2 - d meanmor% 5 e2 s final product to pack, handl d storean e . T7 - vitrification NPH-TO Pools - Products to be vitrified include FP (+ 40 %), clarification - The criticality safety aspect of spent fuel pool storage with , alkalin%) 0 fine3 + e( s solutions (rinsin P evaporatorsF f o g , e storagth e baskets require o verift s burn-ue th y s sooa ps a n alkaline wastes, basic effluents) with an increased volume by e casth s unloadei k d beforan e d fuee th eplaces th i l n i d some 25 to 50 %. These quantities to be processed require that storage basket. three T7 lines work simultaneously. Total Reprocessing waste volume EXTENSION OF BURNUP IN INDIAN PHWRs same th e mas r heavf o Fo s y metal reprocessed hig,a h burn-up fuel would lead to 2 % more waste than for the reference fuel H.D. PURANDARE, P.D. KRISHNANI, K.R. SRINIVASAN MTHM/ d irradiate.Gw 3 3 However o t de compariso ,th e th f o n Bhabha Atomic Research Centre, ratio of waste volumes by the energy produced by the fuel in a reactoe typicaMW 0 r90 lMTUproducin/ h ,Gw show3 23 gs thar fo t Trombay, Bombay, referencR thLW e e fuel, ther e 30. wastf ear o 5 1 e packager pe d India Gwh while for a high burn-up fuel there are 22.9 1 of waste package per Gwh. Therefore for an equal quantity of energy Abstract produced, the high burn-up fuel leads to 23 % less waste volume. e IndiaTh n nuclear power programm bases i e d upon firs e PHWRth tn s i phas e in which plutonium wil producede lb secone th n dI .phase , this plutonium gammad Betan a s activitie hullr fo d sends an s activitiee ar s woule fas th usee t db n i breederd enhanco t s e fissile materials i.e. t stilbu % l 0 remai10 d n an belo increasine % th 0 w2 y b g Plutonium and also generate U2î3 by irradiating thorium. In the guaranteed parameters. supplementary shieldings would be third stage, Th-U2"reactors wil usee producinr lb fo d g power. necessar package th r fo ey handling. Presently many PHWRs of medium size are either operating or are under Technological wastes : non alpha as well as alpha wastes still construction. An effort is being made to extend the burn up of these remain belo guaranteee wth d parameter eacr fo sh category. PHWR casn si e therdelaa s commercian i ei y l operatio fase th t f no breeder reactors. Because of the fast reactor programme, a fuel reprocessing facility is available. This assures availability of plutonium. Hence, recycling plutonium becomes a natural choice for extending PHWRsf buro p nu . It was decided that the bum up extension programme should be implemented in the present reactors without any change in the basic hardware of the reactor. A number of lattices with MOX enrichment in natural uranium or depleted uranium were analysed and two of these were chosen for detailed studies. It has been established that a lattice with 0'4 wt per cent MOX naturan i l uraniucentrae th n i ml seve nineteene th rod f sclusted o nro r and natural uraniuoutee th rn i mtwelv e rods would increas average eth e reactoexie th t f buro r p nfrou presene th m 0 MWd/tt 00 valu 7 o et f eo MWD/te0 1030 . e transitioTh n fropresene th m t natural uranium reacto X reactoMO o t r r would be feasible without any power derating at any stage. The control and safety characteristics would almost remain unchanged with the MOX fuel. 1. INTRODUCTION Indian nuclear power programm bases ei d upon Pressurised Heavy Water Reactors (FHWRs) using natural uraniue fuel in the first phase. The plutonium produce thesn i d e reactors woul usee b d o fue t e dfas lth t breeder reactors in the second phase. These fast breeder reactors would also l)e used cn to irradiate thorium and to enhance the fissile »aterial in the form of CJ1 either plutonium or U .In the third stage, Th-U reactors would be the e papeTh r qlve « sbrie f descriptio e PWJRth ,f o discussen e sth 00 sourc f nucleao e r power production e thre th e basi eTh .r stagfo s e programme problems associated with extension of burn up and gives results of some is the moderate uranium resources and abundant thorium resources in the studies made for medium sized Indian PHWRs. country. A design has been standardised for a small size PHWR in the 200-230 NATURA. 2 L URANIUM FUELLED PHWRs e rang KW d fivan e e reactor f thio s s operatingtype ar s ebee ha ne ,on e reactoTh r consists primaril horizontaa f o y l cylindrical vessel commissioned e morrecentl on poises ei d commisioninr an dfo y g soon. Five called calandria which is penetrated by a large number of fuel channels reactors of this catagory are in various stages of construction. The design arranged in a square lattice. The fuel channels consist of a calandria tube e PHWs HW alsi R 0 o 50 finalisenuabewor e a a oth fn X o suc f d ro an hd and an inner pressure tube through which the fuel bundles are moved by the reactors would «tart soon. on-power fuelling machine. The pressurized hot heavy water coolant flows o (BoilinInditw s aha g Water Reactors(B»Rs eace vintagf HW ho 0 20 ef )o throug e pressurth h e tube cooe .Th l moderato containes ri d w withilo e nth type but does not have a plan to extend the present 20,000 MWD/Te burn up of pressure calandria e controTh . shut-ofd lan deviced fro verticae sar d lan these reactors, in 1982, a feasibility study of using plutoniun enriched contained within the low pressure moderator. Table-1 gives various details fuen thesi l e BWRs carriewa s d futuren outneee I .th d f arises,i , these unitse KW 0 .50 fod ran bot0 h22 efforts coul revivede db efforn .A exteno t PHWRn e buri dth p sn u presently e fuellinTh g schem s on-powei e r bidirectional multibundle shift is being Bade and has to be viewed in the context of possible delay in scheme. For the natural uranium fuelled reactor, 8 bundle shift has been establishing eonnercial fast breeder reactors of the second phase. e decisioa Ath seques o t havo lnt efasa t reactor programmee ,th reprocessing facility for the fuel discharged from THWRs is available. This TABLE-1 ensures availabilit f o plutoniuy r recyclingfo n . Hence recyclinf o g Design dat Indiar afo n PHWRs plutonium .seems to be the natural choice for extending the burn up in PHWRs. 1: Electric power MHe 235 500 contex e presene th th n f tI o t three stage programme, recyclin f plutoniugo m 2: Thermal power to coolant KWth 756 1730 3J0 78 Calandri 0 60 a vessel diamete S rCB in PHWRs would be only a short ter« measure mainly to improve resource 4'. Number of fuel channels 306 392 5; s cm Lattice pitch22.8 6 28.6 utilisatio importann a d nan t consideratio thin ni s respec thas ti t plutonium 6. number of bundles per channel 12 13 7i Maximum channel power HWth 3.2 5.5 should be available for the fast reactor programme when required and the 1 2 6 . 8 Numbe adjustef ro r rods 4 9; Numbe 2 shim/controf ro l rods quality of the second stage plutonium should not be seriously impaired. 101 Zone control compartments - 14 11. Number of shut of rods H 28 e problemTh s arising while extending LWRf d PHHRburo an sp e u ns ar . Numbe12 verticaf ro l liquid poison tubes(SSS2 1 ) 13; Numbe horizontaf ro l poison injectio6 n tube - s managemens a similar fa s a rpowef o t r distributio concerneds i n . chosen. The core is divided into two or three radial zones to achieve radial power flattening in equilibria» condition and in each zone a discharge burn choseo s a i unp tha designee tth d total powe obtaines e limiti r th d n so dan the maxinuB channel powers and the Baxinura bundle power are satisfied. e LWRTh s havproblee th e controllinf no excese th g s reactivity thas ti built in to compensate for reactivity depletion when the reactor operates at power e PHtfRTh . , usin n poweo g r refuelling onls ,ha snaiy a l amounf to excess reactivity present in the reactor at any ti«e in the equilibrium condition. The only ti«e the problem of controlling excess reactivity is PHWa face whes n reactoe i Ri dnth initias rha l fuel loadin fresf go h fuel. This excess reactivity is controlled with poison in the moderator which is removed gradually e radiaTh . l power flattenin achieves i g Bakiny db ga judicious e numbechoicth locationd f ro an e eithef so r fuel bundles with lowe U r conten235 r o thoriut m bundles powen .O r fuellin alss gi o useo t d 0.00 3.UU ii.UO <).U U O ? I O15.O U naintain proper power shape. Hur n up 'UWU/lel Positions of various devices are chosen based on the neutron flux FIG. 1. Contribution to energy from U-235 and plutonium vs bumup. distributio e fuellinth r fo gn scheme auch tha devicee tth s give desired reactivity worths. discharge t 700a d 0 «WD/T plutoniue e th bur, nup m contributio energo nt s i y about 40 percent. Further burn up extension of the fuel and possible in situ burn up of plutonium Bay not be possible in natural uranium fuelled PHWRs 3. CHARACTIRIST1CS OF PLUTONIUM dulimitatioo et reactivityf no . When plutoniu introduces i m reactoe th n i dr A reactor operating with fuel whic contain, U-23s it h ha n 8i s varying in the fora of mixed oxide of uranium and plutoniua (MOX), with a view to amount f o plutonius m isotope different a s t irradiations. These plutoniua decrease requireaent of uraniua, some problems nay be encountered because of isotopes contribute to the reactivity of the reactor and its energy the particular characteristic plutoniuf so m fuel. sitn i uproductiof o f plutoniumburo y p nu wa y b n, differently accordino gt the fuel irradiation. Fig.1 gives the percentage of energy production by U- 3.1 General Characteristic theid san r Effects 235 and plutonium for the case of natural uraniua fuel with nineteen rod s knowi It n tha fissioe tth n cross sectio botf no h Pu-23 Pu-24a 9an 1 Öl CO cluster used in 220 MWe PHWRs. It can be seen that by the tiae the fuel is are higher than for U-235; so are the capture cross sections. At low CD TABLE-2 3.1.2 Power Peaking O Microscopic Fission and capture Cross Sections( barns) at There is a significant difference in the nuclear properties of a 2200 m/s neutron speed for Uranium and Plutonium Isotopes lattice with natural uraniun and one with plutonium enrichment. Fig.2 shows Isotope Fission Capture Beta 0-235 579.0 100.0 0.0065 variatio f thermao n l group fission cross sectio functioa s na burf no p nu U-238 - 3.0 0.0173 Pu-239 741.0 267.0 0.0020 for natural typeo uraniutw plutoniu f so d man m bearing lattice nineteef so n Pu-240 - 290.0 - Pu-241 1009.0 368.0 0.0049 rod PHWR clusters which we call HOX-7 and HOX-7/12, to be described later. Vu-242 - 19.0 - One can observe a sharper decrease in the cross section in the MOX fuel than in the natural uranium fuel for the same change in burn up. Because of this reason, existence in close proxiaity with each other of HOX fuel bundles thernunbea , energie e ev vicinit e fissioef ar 0 rth o 1. n si j yno resonances of Pu-239 and Pu-241 and absorption resonances for Pu-239, Pu-240, and Pu- 241. Prominen fissioe th Pu-23r e tfo nar absorptiod v resonanc9an e 3 0. nt ea resonance at 1.0 ev for Pu-240. Added to this is the fact that the isotopic compositio plutoniuf o n m undergoes continuous change. These facts resuln ti a highly complex contributio f plutoniuo n o reactivitt » d powean y r productio sofe th t n nspectrui PHWRsf no . The following general effects can be observed in LVRs/PHWRs while using fuel enriched with plutonium. 3.1.1 Reactivity Wort Absorbef ho r Elements The spectrun in lattices with plutonium is hardened and because of higher thermal absorption cross section of plutoniu« isotopes and occurence of absorption resonances, the other absorbing eleoents close to the plutoniu« bearing fuel have to compete with it; also in the vicinity of fuel °0.üü 00 00 with plutonium, the flux would be depressed. This would result in the G. 00 B. 00 10. üü reactivit yabsorbee th wort f o rh elenents being lowe case r th f tha eo n i n p u (GWD/)c n r u b ) natural uraniun fuel. FI G2 Therma l fission cross-sectio naturad an X bumus ln v uraniu MO r pfo m with significant difference in irradiation or of natural uranium bundles and developnent efforts would be possible but major design changes would be MOX bundles would lead to power peaking . ruled out e .alternativTh e approac optimiso t s i hreactoe eth r desigr fo n the new fuel which «ay involve redesigning of the lattice, reassessment of control requirements and redesigning of the control system etc. This 3.1.3 Reactivity Coefficients e hardeneTh d spectru d highean n r resonance absorptio plutoniun ni n approach Tnay involve an intensive development programme. Presently, we are not taking this approach. isotopes result in reduction of reactivity coefficients of plutonium bearing In the approach chosen, since the fuel cluster and the coolant channel fuel in comparison with natural uranium fuel. The magnitude of the fuel and design is not to change significantly, the limits on the power that a fuel moderator temperature coefficients as well as the coolant void coefficient rod in the cluster and each of the coolant channels can produce is would be reduced. The latter has a bearing on the safety of the reactor and can be suitably used. finalised. Als e location controe onumbee th th th f d ro l san elemente th d san fuel handling syste» are finalised. The burn up extension would have to be planned within these contraints. 3.1.4 Control The e fueclustee d powelimitth th ro l n n i ro rs (typically givey nb The fraction of delayed neutrons fro» Bu-239 is snaller than for U-235 integral k-detheta) d channean , l powers would mea fuew nne l e thath t and that for Pu-241 lies between the two. The effective delayed neutron configuration is to be so optimised that the ensuing local and global power fraction will depend upon the percentage of U-235, U-238,Pu-239 and Pu-241 distributions are very close to the earlier one, if not better. The power contributioe fuee welth is th lna s la U-23f no 8 fissions. Howeve woult ri d distributio e e optimiseb clusteth n withi e n b ca i rlimine o t dnth f to be smaller than for the case where all fissions are fron U-235. Such a integral k-detheta even witX fueMO y hresortinb l o differentiat g l situation exists to some extent in a PHWR in burn up equilibrium condition enrichmen varioun ti bundlese th rod n si . and the control system acts fast enough to take account of the snaller value A fuel cluster for extended burn up would have to withstand high burn delayee th f o d neutron fractio always ni s possible. , higup h power ramp d preferablsan y should have higher operating margins; there should be provision to collect fission gases. These characteristics OPTION. 4 BURR SEXTENSIOFO P NU INDIAN NI N PHWRS could be achieved by having sufficient clad thickness, graphite coating of Since the burn up extension programme is treated as interin, it would fuel pellets and higher number of fuel rods in the cluster. For the burn up be carried out, to start with, in the present nediua sized reactors for extension that is envisaged right now, it is felt that the present whice desigth halreads i n y standardised. Hence plutoniu» recycling would collapsible cladding «ay not pose any problems. A number of fuel bundles involve mininum deviation in the basic design of the fuel cluster, the with graphite coating of pellets and a cluster of 22 instead of usual 19 reactor e controhardwarth d lan esystem s etc sucn I .approachn ha , snail «od beine sar g irradiate experiencet NAPP-n di ge o t 1 . The licit on the power in each channel according to the coolant flow TABL3 E- in that channel, fixe y orificingb d y stilna violatee ,lb d unlese sth Lattice Calculations to Analyse Various Options plutoniun enrichmen n thai s snaii t t bu burcasle woulp th neu t dno Central Outer k-infinity Max.Bndl. FUEL 7 12 Power REMARKS increase significantly. In case high burn up increase is to be achieved, the Rods Rods KW. global power distribution in the reactor operating with plutonium fuel H.H. H.tt. U.U. 1.1154 420 - should be close to what is available.with natural uranium fuel. This .would HOX-7/12 0.4 MOX 0.2 MOX 1.2373 441 - also ensure reactivite thath t y variouworthe th f so s control devicee sar 0.4 P.C.MOX-7 0.4 BOX H.O. 1.1829 384 - t reducedno . Suc a hpowe r distribution woul possible db e onl propey yb r O.S P.C.MOX-7 0.6 «OX N.I). 1.2074 362 - fuel Management strategies because the peaks in the power distribution MOX-71« D.O. 0.9 MOX U.U. 1.1972 372 LOW BURN DP would depend upon whether fuel bundles with nuclear parameters differing HOX-7/12 D.U. 0.9 MOX 0.7 «OX 1.2473 435 BURW LO P NU largely reside in the neighbourhood of each other. In this strategy of burn up extension, we would be closing the uranium maximue th n o m hea te fue producinth ln clusteri d gro flue .Th x distribution cycle. We would also try to gain experienc« to prepare for the Th-U233 in the cluster would decide the maximum power that a bundle was capable of reactore thirth df o s stage with mininu desigf o m e n th change n i w sno producing. Investigations haved ro bee 3 2 n d carriean 2 t 2 wit ou , dh19 present reactor d withouan s t losin e presenth g t reactor performance clusters with plutonium enrichments eithe naturan i r l uraniu depleter mo d Characteristics. A cycle using thorium, Once Through Thoriun (OTT), in which uranium oxide. irradiation of pure thorium bundles would be segragated from irradiation of Aftex analysing different cases, as shown in Table 3, the following MOX fuel bundles s beini , g considered thin .I s cycle, benefi sitn i uf to two fuel lattices were studied in detail to start with, (a) MOX-7 with U f produceo bur p u n d woul obtainee db experiencd an d e with irradiation central seven rods having 0.4 wt.percent plutonium oxide in natural uranium of thoriua would also be obtained, whereas the option of closing the thorium e outeth oxid rd twelvan e e rod f naturaso l uraniu mHOX-7/1) (b oxid d ean 2 cycl exercisee b neet dno d immediately. having central seven rods with 0.4 wt.percent and outer twelve rods 0.2 e nexTh t section give experiencr sou establishinn i e g feasibilita f yo wt.percent plutonium oxide in natural uranium oxide. scheme to extend burn up by recycling MOX in the medium sized reactor with There wer o problenscheme tw eth en i sviz . desig reactoa f no n i r minimum deviation from reactor parameters using natural uranium fuel. equilibrium with MOX fuel and the transition of the present natural uranium SCHEMA . 5 R EXTENDINEFO MEDIUR G FO BUR P U MN SIZED PHWRS X reactorMO e reactoth . o Whilt r e optimising botr casese fo th h,e th , It was decided that the integral k-detheta of 40 watts/en as in the constraints of total power, limits on bundle and channel powers, limitations case of the medium sized PHWR using natural uranium fuel would be the limit on the fuel handling system and the control system were satisfied. 5.1 Equilibrium Reactor 5.2 Transition to BOX Reactor For the MOX-7 reactor, a central region with all natural uraniun fuel The studies carried out show that it is possible to change over fron n outea d r an region with MOX-7 fuel were devised while retaining eight the present natural uraniun equilibria« reactor to a HOX reactor without any bundle shift. Thiessentias wa s avoio lt d peakin centra e powef gth o n ri l derating of the reactor power. This however would require introducing core that otherwise would have occured. The average exit burn up of the natural uranium fuel bundles along with the HOX bundles to avoid axial power reacto rp frou woul no g d700 0 HWD/T 1030o et 0 refuelline MWD/Tth d ean g peakinfirse th tn gi phas transitiof eo firse th tr noptiofo MOX-f no 7 fuel. requirement would cone down fro averagn na 4 fue 7. l f ebundleo o fivst e fuel The transition in the case of MOX-7/12 would require introduction of some bundle eacr sfo h full powe. rday depleted fuel bundles in addition to the natural uranium bundles. e reactoTh r with MOX-7/12 could give exit f 1350buro p n0u HWD/Te in order to avoid introducing natural/depleted uraniun bundles as well without violatio contrainty an f n o combinatio a fou d f an si r o bundltw f no e as for avoiding the central natural uranium fuelled region, higher than 19 refuelling scheae is resorted to. In case the present eight bundle scheme is rods cluster coul helpfule db . continued e maximuth , m channel power liait woul exceedede db . With this The transition to a HOX reactor could also be carried out by loading requiienene th , up exi e valum refuellinf th to t bu f eo g would come downn ,o the reactor with all HOX fuel bundles at the time of replacing the present an average, to 3.8 fuel bundles per day. However, the following development pressure tubes. This situation would require controlling excess reactivity effort woul necessarye db . which woul e higheb d e fresrth h r thanaturafo n l uranium fuel whee th n n averaga r Fo e exi) (1 t f 1350buro p n0u HWD/T maximue eth m burp nu reactor operation starts. Also the core radial flattening would have to be the fuel would go up to about 20000 MWD/Te Fuel development to -withstand optimised with depleted or thorium bundles. But with the experience of this high value woul necessarye b d . controlling such reactivit d achievinan y g radial e flatteninth l al n i g (2) Plenum for fission gas collection would be necessary. earlier natural uroJiium fuelled reactors, this should not pose a problem. (3) In the case of a combination of two and four bundle refuelling scheue e fueth , l bundles would havwithstano t e d power ramps. Graphite lubrication of the fuel Bay solve this problem. 5.3 Safet Controd yan l Characteristics (4) Though the nunber of bundles to be refuelled would be less than s founwa It d thareactivite tth y e controwortth f ho l system, called four per day, on an average, two channels may have to be visited every day. adjuster rod system, would decrease only by about four percent for the MOX-7 The fuel handling system should be geared to take this load efficiently. reactoeighy b d t ran MOX-7/1e percen th r fo t2 reactor. CO CO TABLE - 4 BURNUP EXTENSION PLAN OF BWR FUEL AND Void Reactivit t Differenya t Burs up n IMPACFUES E IT TH LN CYCLTO JAPAEN I N Burnup Void Hectivity(nk) A. TOBA (GWD/T) Natural uranium MOX-7 MOX-72 /1 Tokyo Electric Power Company, Inc., Tokyo, Japan O.O(NoXe) 8.47 7.94 6.57 Abstract 0.0(wite hX 8.94 . 8.26 6.78 4.0 6.52 6.44 Japan I n bur extensiop nu f BWno K fue a beinl l g pursue a stepwis n i d y wa e LJ5. to 45 GWd/t (batch average). Within this burnup range, no major 5.61 6.16 technical issue and no major licensing program will be expected. In case 6.0 6.22 of burnup extension over 45 GWd/t some modifications will be necessary throughout the nuclear fuel cycle to prevent problems on criticality, 8.0 - 5.77 5^21 shielding, etc. 10.0 4.94 5.38 5.62 . 0 BURNUP EXTENSIO FUER NN JAPAI LBW PLA F NO The reactivity addition when all the coolant gets voided is alnost the san r e seebot b e reactorfo e nth hn ca fro s ma s Tabl. 4 e 0.1 Introduction It was also found that the minimum value of beta effective of both the n JapaI n numbe f nucleao r r power plants have been steadily reactors would not be significantly different fro* the value of beta operating in these years and playing quite an important role in effectiv e naturath f o e l uranium reacto t equilibriuma r . electricity supply with fairly high capacity factors. Nuclear power generatio s severanha l advantages compared with other power generatio e prominenth f no e methodton featured an s s 6. CONCLUSIONS shouls lowe ha thae b drt i tpowe r generating cost (total power Pros the studies carried out on the nediun sized Indian PHWRs, generating cost including construction cost, maintenance cost, fuel cost and etc) . Extension of operating cycle length will extension of burn up in this size reactors is feasible with plutonium further improv e economicath e l advantage f nucleao s r power plants obtained froe naturath a l uranium fuelled e PHWRsschemth n I e. that coule b d through higher capacity factor thin i , s s favourabli cas t i e o t e use the fuel which can be usable in the high burnup regions. In started with «ini»un of développent effort, saving of natural uranium fuel additio o thist n , extended burnup result n lowei s r fuel cycle of about 40 percent is possible. The transition to such a reactor is cos e cyclt th eve ef i n lengt s kepi h t constant. Burnup extension possible without power derating. The control and safety character is ics would also e decresleadth o t f snumbe o e f speno r t fuel, whics i h desirable to mitigate the burden on backend of fuel cycle. So it renain almost unchanged. is desirable to introduce high burnup fuel in commercial use as soo s possiblea n n thosI . e background, developemen f higo t h burnup fuel has been pursued. 0.2 Target Burnup 84 85 86 87 88 89 90 91 92 9394 95 96 97 98 99 Extended burnu s significanha p t e economicaimpacth n o t l Batch Average Discharge Exposure aspects of nuclear fuel and earlier introduction is desired as Y YFueJ l Assembly Maximum Exposure mentioned already. But as the target for high burnup goes higher and higher, the task for research and development would become New 8X8 ()- the more difficult. In this connection it appears to be a more effective and steady way to apply the technologies already establishe aftee don r t forworanothedirectioe pu th d o t dran o t n New 8X8 Zr Liner Approval high burnup in a stepwise manner. Our approach on the (Step I) development of high burnup fuel is shown in figureO-1. Design Fabrication Current database on the high burnup experiences in the foreign commercial reactors demonstrates that no particular failures associated witincreasee th h d burnu e see thin ar pi n s Approval burnup range( 52GWD/t) have .W e been pursue developmenr dou t Hig8 h Bur8X p nu (Step fl) [ evelopmen Desig& t n Fabrication Approval ^:-: ~ ACCUMULATION and EVALUATION of DATA Step I Developmenr Desig& t nr mj* Fabrication ESTABLISHMEN f STEPo Tr fo S R BurnuBW 2 p0- Fig .Extensio n Plan HIGH BURNUP TARGET e efforfollowinth n i t g steps. Targe f higto h burnup fuel [DEVELOPMEN HIGf o T H BURNUP TECHNOLOG| Y introductio d theian n r schedul 0-2g e showfi ar e .n i n Stepi Average discharge burnup approx.33GWd/t Maximum assembly burnup 40GWd/t Step U Average discharge burnup approx.39GWd/t EARLY COMMERCIAL INTRODUCTION Maximum assembly burnup 50GWd/t o> Ste pu Averag e discharge burnup approx.45GWd/t UI Fig. 0-1. Steps for High Burn-up Maximum assembly burnup 55GWd/t 0.3 Fuel Design oO> )r LineZ Ne8 r Fuew8X l Hig 8 hFue Bur8X l p nu (Step I Fuel) (Step ÏÏ Fuel) (1)Stepi Fuel Stepi fuel has been used since 1986 in Japan. This fuel utilizes the Zirconium barrier fuel, and enough reliability can beconfirmed through current data Thoug( . h over 8000 fuel assemblies have been introduced in Japanese BWR core, there is no fuel failure yet.) (2)Stepn fuel Upper Tie Plate Upper Tie Plate ( Small Pressure Drop Type) Stepn Fuel is the next stage fuel whose batch average Channe— x Bo l discharge exposure is about 39 GWd/t. Figure 0-3 shows a comparison between Stepi fuel and StepII fuel. In StepII fuel, ferrule type spacer is introduced. Using the ferrule type spacer. ii- t Critical Power Ratio can be improved about 15%. Another structual characteristi StepIf co I fue largs i l e '"' -É center water rod which occupies the four fuel rods region in the bundle. Ferrule Type Spacer Spacer (Egg Crate Type) Because average bundle enrichment of StepH Fuel(3.4%) is Improvement of Critical Power Ratio greater than that of StepJ Fuel, neutron spectrum hardening (Goo OOOOO cause worse changsa th voi f eo r defo coefficien shud tan t down oooooooo OOOOOOOO ooooo ooo OOOOOOOO margin. Therefore, it is necessary to soften neutron spectrum OOOOOOOO OOOO0OOO by increase of H/U ratio. To increase H/U ratio, large center OOO0OOOO oooooo o o OOOOOOOO OOOOOOOO introduceds i wate d rro . OOOOOOOO DOOOOOOQ 100000000* Water Rods Large Center Wated Ro r (3)Stepffl Fuel Increasf eo Japanes utilitieR eBW s have been investigate o typetw d f so H/U Ratio fuel desig steps na n Fuel (figure 0-4). directioe Th f improvementno s towards burnup extensiof no Stepm fuel are basically same. In nuclear design terms, it is necessar increaso t y D ratioH/ e . In Stepn fuel, bundle average enrichment wil e increaselb d to about 4.0%muse w e bigge o tus .S rfuen i wate ld bundlero r n I . Lower Tie Plate Lower Tie Plate Step ru A-type fuel, two central water rods will be introduced Fig. 0-3 Structure of Step II Fuel into the seven-fuel rods region. And in Step m B-type fuel, large water channel wil e introduceb l d inte nine-fueoth l rods region. Most remarkable design change of Stepffl fuel is introduction lattice9 x O 9 f9 lattic.9x e arrangement bring certain merits. One is 20% decrease in average linear heat generation rate and greatea anothes i e r ron flexibilit fuen i y l assembly design. For example, flexibility of Gd fuel rods arrangemnet will n alsincreaseca o e w simplif d ,an y fueenrichmend ro l t sprit design. upper Tie-Plate 1.EFFEC BURNUF TO P EXTENSIO NUCLEAN NO R FUEL CYCLE STRATEGY 1.1 Overview Japan's nuclear power generation has been established technologically compared ,an d very favorabl performancn yi e with Large Center later Channel that of the other countries which have nuclear power plants. In nucleae th r fue ablw lfabricato no et cyclee e th ar l e ,w eal nuclear fuel we need, but as for such critical processes as enrichment and reprocessing we mostly depends on overseas facilities. establiso T entire hth e nuclear fuel cycl Japann ei o ,tw Ferrule Type Spacer companies are planning to construct three key nuclear fuel cycle Ferrule Type Spacer facilities at Rokkasyo-mura in Aomori Prefecture, Nortern Part of Japan. These facilities are a spent fuel reprocessing plant, uraniua m enrichment low-leveplana d tan l radioactive waste storage facility. 1.2 Effect On Front End Strategy Lower Tie-Plate (1)consumptio naturaf no l uranium Loier Tie-Plate To ensure the steadily supply of uranium resources, its sources shoul diversifiede b d ratie ,th developmenf o o import t TypeA TypeB should be increased, and prospective resources (interests) should be acquired whereever possible. 4 Structur0- . Fig f o Stepffe l Fuel Figure 1-1 shows relative comparison of natural uranium O) consumption among high burnup fuels. CO Though higher burnu ptendenca fues lha decreasinf o y n ga amount of natural uranium required, the defferences are not so R Valu large e reaso.Th thif no s decrease results from decreasing batch TJ C size, because these calculation is based on the same operational period, high burnup fuel smala whic s lhha batch siz burnes ei d O •Sp g3 o effectively case th e n .thaI t operational perio extendes i d n i d 01 propotio extendeo t n d burnup n amoun,a naturaf to l uranium required become almost constant. Sinc effece eth burnuf to p extensio consumption no f no natura o largels uraniut ,no s Japan'i m s future strategy wilt lno be affected. Japanese electric power companies have secured some o •- amount of U-jOg by long-term contract with overseas mining companies. This assured quantity of uranium will meet the O- XJ O \ O domesti ctime demanth e r beingfo d . However, give depressee nth d o state of exploration and development for new resources o 3 worldwide,along with the inevitably long lead time for uranium T3 t» development,the global supply-demand situatio uraniuf no y mma get tight. So shoule ,w encouragee db ensuro t d e long-term, stable supply 0 E) of natural uranium. CO co O o H W (2)Enrichment T) CJ \ -I «-*• ö In Japan, uranium enrichment services has been consigned to united State Franced e othesan th n r.O expectehands i t ,i d that smooth and steady operation of domestic enrichment facility will be promoted by utilizing the already established centrifugal separation method. A gradual expansion of domestic enrichment en in facilities is expected. c O ————————————————————————————————————————————————— CD W Figur show2 1- e s relative compariso requiremenU SW f no t among *-t Q. ^d CD high burnup fuels. differenco Thern s i e e among three typf eo 3 \ ————————————————————————————————————————————————— CO fuels. CD r-r a 3 (3)Fuel Fabrication Figure 1-3 shows relative comparison of fuel fabrication services among three type of fuels. As the amount of fuel OJ 1 0 . 1 3 0 . 1 3 0 0 . 1 1.0 CD 0. 5 i- o a: 0 Fuel Typ » - e StepI STEPïï STEPm Dis.Exp. -» 33GWd/t 39GWd/t 45GWd/t Fig. 1-2, Relative Comparison of SWU Requi rement 1. 00 *" J. . U 3 0. 85 « 0. 7 6 O) 5 . 0 > — 03 0) (ü 0 Fuel Type -> StepI STEPïï STEPM Dis.Exp » - . 33GWd/t 39GWd/t 45GWd/t Fig. 1-3. Relative Comparison of Fablication 69 -4 fabrication servic fuer epe l bundl almoss ei t same, burnup O extension contributes to decreasing fuel fabrication services for one cycle operation ( usually 12 month in Japan ). R t i o Current domestic fuel fabrication capacit enougs i y o t h M- m satisfy domestic demand. Moreover, the lead time required to build fuel fabrication facilitie relativels i s y short woult i , d •n 01 o be easy to increase the capacity in a short period for expanded w ~r~ T~ power generation. 1.3 Effec Bacn tStrategO d kEn y £ co O r* (1)Reprocessing o. u o \ o In the case that Japanese utilities use high burnup fuel for 1 year's operation, fuel batch size decreas expectede b n eca . Figur show4 1- e s relative compariso amoune th f spenf to no t fuel after 1 year cycle operation. For example, spent fuel generation O o will decrease about 25% by using Stepffl fuel, when it is compared 3 with using Stepi fuel. •o Although Japanese utilities have made reprocessing contracts 39GWd/ t o> with BNF COGEMd Lan thesn i A e years, spent fue expectes li e b o dt H O W reprocessed domestically in future. First commercial reprocessing oo plant(capacity:800t/y) will start operation around middle of Öl 1990' . H Before reprocessing, spent fuel will be stored properly- Requirement studiegoine s b si o gimplemented t dan thao de s tth •o spent fuel can be stored in dry and in high density. u Sï to 2.1 Increasing Reactivit Core Associated Th eAn n I y d Problems O H With Fuel Management And Reactivity Control ^ W a. T) \ 03 o extenT d fuel discharge burnup generalls i t i , y requireo t d - H increase enrichment. There is a tendency that simply increasing enrichment cannot gives us most of the merit, because of spectral hardenin gconsequence whicth s i h f increaseeo d thermal neutron absorption. Increased enrichment also leads to increase of reactivity defferenc powed an e r mismatching between fuel bundles. z S c •H l D O n(N \ •s Harden Increase ——>• ——*• Increase AMCPR during Void Coefficient Neutron Spectrum Transien Increas( t e Operating Limit MCPR) N f \ Increase Reactivity Reductio f Shutdowno n —> • Difference between Cold Margin Condition and Operational Condition Reduction of Thermal Margin Increase Power Mismatch Fig.2-1 Bur Extensiop nu n Charaoterisity (Nuclear Aspects) 71 The main effects of the burnup extension in the nuclear propertderivel al e dar y from thes o characteristicsetw s .A typicaly illustrated in Figure2-l, spectral hardening and Parameter : Enrichment increased power mismatching leads to three effects listed below. -Increased absolut evoie valuth d f reactiviteo y coefficient -Increased cold-hot reactivity swing -Increased power peaking 1.6 12H/0 consequence th s A thesf eo e effects changeo core ,tw th e f so property listed below occur. 6»/0 -Decreased thermal margin -Decreased cold shut down margin In the following sentence, the countermeasures of these 1.4 changes are described. 3W/0 Optimizatio) (1 watef no uraniuo rt m ratio (Optimizatiof no lattice design with increased enrichment). Firs shoule tw d considedischarge th w rho e burnup will change as the increased enrichment. With these effects, we must conside change e initiarth th f eo l K-infinit change th d f eo an y 1.2 K-infinity with burnup. Figure2-2 show maichange e th sth n f parametereo s which consis K-infinityf to , using wate uraniuo rt m ratia s a o parameter. This figure also show parameterchange e sth th f eo s with increased enrichment:the thermal neutron utilization factof r increases with increased burnup. The slowing down probability p slightly decreases, because of spectral hardening. As the total effect of these parameter changes K-infinity changes as illustrated dotted line in this figure, thepeak value point change o right s t sido highe(t e r wate uraniuo rt m ratio).This H/U (Atomic Ratio) mean conditiose th tha n i t f usino n g same wate o uraniurt m ratio, Fig.2-s k-inv U fH/ 3 difficuls ii t obtaio t maximue th n m gain byincreased enrichment 1.6 Parameter : Enrichment , H/U 160 ~ 12H/0 H/U- 6 140 1.4 12H/0 H/U-4 120 100 ~ tfl 80 - O Sf 60 ~ 40 - 20 ~ 6 9 12 Enrichment (wt/X) Fig.2-5 Average Discharge Exposure vs Enrichment Exposure (GWd/l) Fig.2-4 K-inf vs Exposure CO because the point will be far from peak point and K-infinity increase will be very small. exampler Fo , Figure2-3 show e relatiosth n between K-infinity d enrichmenan parametea s ta f hydrogero uraniuo nt m number density ratio. ' fi But spectral hardening causes not only bad effect, but also recovery effect of K-infinity in the high burnup region based on S the increase generatiou dP d highenan r conversion ratio. ça ft To determine which effect is dominant, the burnup calaulation was made (See Figure2-4). Based on this result, the M average discharge burnup was caluculated by linear reactivity I model (See Figure2-5 shows A )thi. n i n s Figure achievo t , e V higher average burnup favorabls i t i , havo et e highe U ratirH/ o Dl in spit f recovereo y effec f K-infinityto vers i yt i o .S n important to adapt optimized fuel lattice design in the view W U ratiopoinH/ f .to h H- As mentioned above, higher H/U ratio is essential in the view ay poin f reactivityto alss i ot .I effectiv countermeasura s ea f eo a>B 3 increased peaking and decreased cold shut down margin. r* < ) Powe(2 r peaking factor •o o desigT conviniencores R i nBW t ,i divido t e power a P> distribution into some factors r thi.Fo s purpose three power *• H- peaking factor definee sar describes da d below. 3 Radial peaking factor: peak bundle powe averago rt e bundle power ratio Axial peaking factor : axially maximum planer power to average power ratio &ru Local peaking factor : maximum fuel rod power to average fuel d powero r bundl e ratith n i eo cross section consequence th s A burnuf eo p extension, radial peaking factor increasee tendb o t s d because increaseth f eo d reactivity defference between fuel bundles exampler .Fo , accordino t g increased average discharge burnup, from 30GWd/t to 45GWd/t, Cold Uncontrolled k the radial peaking factor is increased 3% in batch average(the eff> '•' same batch sizsamd ean e fuel lattice configuratio e assumed)ar n . Local peaking factor tend to be increased, because of thermal neutron defference between peripheral region and central Hot-Cold region in the fuel bundle and increased number of the gadorinia Swing rod suppreso st s large excess reactivity(see Figure2-6). consequence th s A thesf eo e effect margie ,th o thermat n l Cold limi decreaseds i tnecessars i t i o adequatel.o S yt y control Blade axial peaking adequate factous d ran e fuel lattice configuration Worth to suppress large local peaking factor: as for BWR the techniqe to control axial peaking facto bees rha n establishe meany db f so axially zoned reactivit thie us s n techniquca y e corew d o ,an et Operational Condition generate enough thermal margin. As for the local peaking factor, keff= 1.0^ n suppresca locatin y b e w t centrae si th gn i watel d regiorro f no (Standard point) the bundle. Shut Down Margin (3) Improvement of void reactivity coefficent e voiTh d coefficen defines ti describes da d below. d Keff Cold Keff (Vchannen i : l void fraction) ef f< 0 .99 (One stuck control The void reactivity coefficent is an important factor to rod condition) dominate transient even accidend tan BWRt a t . Because eth Maximum absolute value of coefficent is more than ten times larger than Rod the other reactivity coefficent which is common to other type of Worth reactors, doppler coefficent, moderator temparature coefficient and etc. In case of increased enrichment without modification of Cold Controlled fuel lattice configuration, the sensitivity of K-infinity on water to uranium ratio, the absolute value of void coefficient, ————— Base Condition tends to be increased. High Burnup Condition The increased absolute value of void coefficent tends to affect some results of transient events and leads to some Fig 2-7 Degradation of cold shut down margin Ol difficulties such as decreased thermal margins y higb h burnup -J alss i ot i desirabl o S modifo t e y fuel lattice configuration —Inclease initiaH e th l f pressureo d optimizatioan e f o n O) to achieve adequate wate uraniuo rt m ratio accordin o increaset g d pellet-cladding diametral gap enrichment froe stanth m d poin f voito d reactivity coefficent. These lea o improvet d d thermal transmission characteristics. ) Improvemen(4 colf to d shut down margin s increasFga P eeffecd presba d n thermao tsan l performancf eo fuel rod. The accumlation of FP gas reduce the thermal Figure2-7 conceptually shows the relations of reactivities transmission rate, so the pellet temperature gets higher. As the relate colo t d d shu treactivite dowth R n BW margin f r yo fo s .A burnu s incleasedi p e initiaH . l pressur e fueth lf o eshoul e b d the core becomes maximum value on the cold condition(@in increased. Figure2-7), because on the operational condition(®in Figure2-7), there is a negative reactivity feedback. This reactivity —Optimizatio f claddinno g thickness defference between cold conditio operationat ho d nan l condition is called cold to hot reactivity swing. If all control rods are To extend burnup with sufficient design margin, the inserted in the core, the core reactivity decreased by worth of optimizatio f claddinno g thicknes desirables i s . control rods. Whe controe nth whicd maximus lro hha m worts hi withdrawe thit da s condition core th Kefe e f ,th becomefo e sth Ferrul— e type spacer value at ® . The reactivity defference between (5 and @ becomes cold shut down margin. Ferrule type spacer produces better CPR performance and casn I f increaseeo d burnup without modificatio fuef no l maintains thermal margins. bundle configuration spectral hardening leads to increased cold t swingho o .t Thus cold shut down margi decreases ni d showa s na —High-Flow upper tie plate dotted line in figure2-7. To avoid this effect, the increase of water to uranium ratio Reduce two-phase bundle pressure drop and maintains thermal- with modificatio fuee th l f latticno consideres ei a s a d hydraulic stability margins. countermeasure. 2.2 Optimization Of Fuel Assemblies And Fuel Element Design To (2) Burnup extension beyond 45 GWd/t Overcome Technical Limits R*D program was initiated to develop new fuel materials in (1) Burnup extension within 45 GWd/t order to explore the possibility for burnup levels up to about These designs described below will be introduced. 70 GWd/t batch average. Increas— f fued plenuo e ro l m volume Improved pellet materials and zirconium alloys are now being irradiate e Haldeth n ni d reacto higa d h an rpowe r density This lead o reductiot s e internath f o n l pressure e fueth ls .A commercial reactor in USA. A part of the specimens has been burnu s increasedi p s alsi e productio os th , ga increased P F f o n . retrieved and examined in GE and NFD hot labratories. From 1990 This increases the fuel rod pressure is so high, fuel cladding is the Halde R corrosioBW n n loop experiment n whici , h EPRs i I lifted offd effecba . a n Thifue o ts i ls mechanical reliability. partially participating, starte o investigatt d e effectth e f o s reactor coolant water chemistry on zirconium alloys and to month assumeds ,i maximue ,th m valu s lesi e thin sI thas. 5% n accelerat e selectio th ee improve th f o n d zirconium alloys. case, the impacts on front end become as described below. After selecting new improved materials, the necessity of safety-related experiments A simulationsucRI s LOCd a h an A s will (1)Fuel fabrication and reconversion be considere e irradiatioth d an d f LeaAssemble o n Us d y incorporating improved materials wil plannee b l detailn i d . From the stand point of criticallty control, the maximum These RSD activities are also useful for the burnup extension batch quantity must be decreased according to the enrichment. But level less tha GWd/0 7 nd bac'ktan f itable. the modifications of the current facilities will not be needed, so the impact is small. 2.3 Licensing Problems (2)Uraniu d fresan m h fuel transportation licensingr fo s A iteme ,th s described below shoule db Current containers can be basically applicable for the considered. transportation, though the maximum capacity should be limited in some cases. -The maximum value of average discharge burnup is described in E/P submitals. In case of burnup extension, Ministry of International Trad Industry,MITI,and ean d Atomic safety Committee 3.2 Burnup extension beyond 45GWd/t review the submitals even if it is a small number. The design method usee b thin i do s t s review were already reviewe thesey db s The maximum value of enrichment may exceed 5%. In this case, organization summerized san reporte th n i d. the impacts on front end should be investigated as described below. fuee th l r fabricatiofo s -A n facilities applicable ,th e enrichment range of the guide line is limited to less than 5%, (l)Fuel fabricatio reconversiod nan n case ith ne that more tha enrichmen% 5 n t uraniu necessarys i m , this guide line woul e reviseddb . The applicable enrichment guidelin e rangth f eo r safet fo e y analysis of fuel fabrication facilities in Japan is limited to -The lengt operationaf ho 3 l1 cyclo t s limitew i e la y db les casn sI thae. tha5% n t more tha enrichmen% 5 n t uranius i m month burnue ,th p extension doesn't directly lea extendeo t d d needed, the revise of this guideline is needed. cycle operation. The change f criticalitso y controe facilitieth n i l s wile b l needed don.e decrease b mean y Theth maximue b n th ca f yso f eo m . Impac3 f nuclean o fronto d ten r fuel cycle batch quantity and modifications of the facilities, though the efficiency wil decreasede b l . Figur show1 facilitie3- e sth s 3.1 Burnup Extension Within 45GWd/t which suppose affectede b o t d . Froe stanth m d poin f exposurto e control e increas,th f eo The maximum value of enrichment depends on the assumed exposure dose wil e relativelb l y smaller majoe th ro ,s operational cycle length t eve bu ,f longei n r cycle,8 1 suc s a h modificatio facilitiee th f o n needede sb wil t no l. --J (scrap uranium iUO., Powder] — )foO2 Pellet! Heatin, UP g b * * Powder Arrival Loading Hydrosis (Tank Diameter) Solution (Tank Wall ThicknessX-i Solution* (Tank Diameter) Storage* Rod Drying (Can Spacing) PH Control* Solvent Extraction (Tank Diameter) UEP Welding Precipitation (Machine Shielding) Precipitation i Mixing* |Fuel Rod] Scrap Mixing Machine \>_ Separation (Diameter) Uranium (Hall Thickness) Separation /* From Storage* Calcination all process Pelletizini g d Ro r ITray Spacing] Drying Reduction (Diameter) Sintering* Assembling rPellet Board] Pulverization (Diameter) Reduction [Height 1 (Fuel Bundle] Mixing*(Machine) Grinding* i Powder Dust i I . \Tank Wall Thickness) Storage* Storage [Storage Volume! (Bundl eSpacin, g 1 Drying Shipment * Process necessar o considet y r modification. Storage* [Pellet Tray Spacing] Facilit) ( y modification wil needede lb . i— —— r— Drain 1 Batch control will be needed. * Process necessaro t y Storage* consider modification (Tank Wall Thickness) ) Facilit( y modification I will be needed Drain,Treatment Batc1 I h control wil e needeb l d Flg3-l-2 Conversion Process and Facilities necessary o considet r modification- Fig.3-1-1 Fuel fablication proces facilitied an s s necessary to consider modification As for ventilation, drain and solid waste control, the impac f increaseo t d enrichment also wil majoe smalle lb th ro s , modificatio facilitiee th f no needede sb wilt no l. (2)Uraniu d fresan m h fuel transportation According to IAEA Safety Standards published in 1985, the package which contains a few Kg of uranium with over 5% enrichment will be categorized into B-Type. So the current package for fresh fuel transportation will not be applicable. The capacity of the current package for U02 powder will be limiteo smalto lo t dvalu transporo t e t efficiently. As for UF, transportation, current cylinder will not be applicabl regulatioe th y eb DOE,ORf no Rev61 O65 . Therefore cylinder wil needede lb . . Impac4 Bacn td O kEn According to burnup extension, the increase of accumulated activities, decayheat generatio numbed nan neutrof ro n generation ocurr. The calculated examples of these effects are shown in figure 4-1^3 as a function of cooling time. The calculated example of FP generation is shown in figure 4-4. 4.1 Burnup Extension Within 45 GWd/t (1)Spent fuel transportation 5 1 0 Cooling Time (Year) As for the critlcality, the current casks will be applicable with modificatio e heath t r baskef caske o nfo th s .n A i t removal,they will be applicable with extended cooling time, 1 FigActivit4- . s Coolinv y g Time almost 6 month. From the view point of shielding, as for the gamma-ray they will be applicable with extended cooling time. As neutrone th r wilt fo i ,e require lb optimizo t d e numbeth e f ro -vl (O assemblie casa d shieldin r an k fo s g structure. XI 0s CD 40 O 3.0 .*! i o c o oV 3 V Z CJ 1.0 XI S 3 Z 4 SGWD/T Cooling Time (Year) Fig. 4-2 Heat Generation vs Cooling Ti S l 0 Cooling Time (Year) Fig. 4-3 Number of Neutron Generation vs Cooling Time 0 5 l 4.2 Burnup Extention Beyon GWd/5 4 d t (l)Spent fuel transportation n H From the view point of criticality, heat removal and \ 6C shielding supposes i t i , d thamajoe tth r modification f casso k desig d longenan r cooling timneedede ar e . C O l 0 0 Table 4-1 Specification of JNFS Acceptable Spent Fuel and Extended Burnup Fuel ^^-^_ JNFS Acceptable Extended Burnup Spent Fuel Fuel 3 T3 SO O Bundle Average u Initial Enrichment (w/o) D, 5 5 4. < before Irradiation C O Maximum Allowable Bundle Average (w/o)3.5 Enrichmen Spenf to t Fuel J_ 0 50 100 o s i Average Burnup v(GWd/t) Burnup (GWd/t) par Day ' 45 < 45 Fig. 4-4 Fission Products Generation v s Burnup * Bundle Maximum Burnup (GWd/t) 55 55(60) (2)Reprocessing facilities Cooling Time — before Shearing («ear) 4 e JNFTh S reprocessing facilit designes i y e ablb o t eo t d reprocess burnup within 45 GWd/t. Table 4-1 shows the basic oo specification of the fuel which JNFS can accept. *Desig) ( n Burnup GO (2iReprocessing facilities ro Cfl ?o c/) 3 t Hr Z, 3 1 1 Figur show5 conceptua4- e sth l illustratio f reprocessinno g I.2- S process floanticipated wan d problems with extended burnups .A V E-U shown in this figure, these problems described below should be s S s =^ considere connection i d n with extended burnup. CD S 3 a o O ä M Ä »-t» • — o s •Ä • i S " O t r- J t» -The thicknes f shieldinso g wall mus increasee tb n i d "•2 S s s i7 r-f (» O " &a s proportion to the increase of neutron generation. jps. (M g- ^ s -r O u CD ta •3 o S O S. a -The increas f filterineo g ability capacit facilitiee th f yo s ca. i" s e-s T wil needee lb copo t d e wite increasth h nuclidef eo s which ^ SO CP HH 0 !-• O^ » CD i» C ï te n> n SS 8£1 CL r-t need release control to the environment. ? " £ § saD M» D t-ö 1 ScÏ it § S S" -Preventio blockagf no liquif eo dsludgo t flo e wedu rt- (D 00 r fD 0. < depositio increasy nb clearinf eo g capacity wil needede lb . V CD M c? r-t »-t, -Shortening of extraction time or the increase of frequencies 5 S" s S I Da § of solvent regeneration will be needed, because of activity M il^ffi a « CD CO increase in the solvent resulting from FP. g g t—t h-* *o o. c5 1 i h-h CD -The increase of disposal capacity of radioactive waste O) disposal facilities and storage capacity of solidified high CQ § level waste will be needed as the consequence of increased T) a; •n M t r C » 2 S ••t FP generation. o o 1— < , -- t-« - » 0S. «l 0- § ? PO 2* $ T3 While iteme ,th s listed below shoul consideree b d n i d « PO r+ •o S M h-+» H-h f nff connection with modifications of fuel design : additive 3 i O g ö — • o t r e _ •< * s l?s D C r r c u g- g i , a B )> * contained pellet, new cladding materials and new bundle M o C 1 0 3 i Q. 3 0 Q O 0 ï 3 rï CB t-t- configurations. » » 3 iî as » "SS- _ t3 O ^ r 1 r Conversio n Conversio n Plutoniu m Treatmen t -Impact of pellet material on solubility and insoluble 3 M.S. UraniDu a W S n t- residue. t— ' g 1- t"" =•] •^ g 8 » -Possibility of increased cutting powder of cladding material and deformatio f cuttino n g face -Applicabilit f cuttinyo g machin o bundlt e e configuration. HIGH BURNUPS FOR WATER REACTOR Government also decided to postpone a decision on whether to proceed with three of the FUELS - A UK PERSPECTIVE planned PWRs until after a major review of the nuclear industry to be conducted in 1994 However, the Government confirmed that the first PWR in the series, Sizewell 'B', which was already under construction would proceed to completion K HESKETH British Nuclear Fuels pic, Thus we have the current position in which only a single nuclear power station is under Salwick, Preston, construction in the UK, at Sizewell 'B' and plans for other stations are on hold pending a United Kingdom major review Following the review, it will be decided whether Nuclear Electric should proceed with duplicates of Sizewell 'B, opt for alternative PWR designs or not proceed Abstract with any new nuclear plants An encouraging recent development is that British Nuclear Fuels is actively exploring the possibility of constructing one or two PWRs at its Chapelcross and Sellafield sites The first phase of this study has recently been e presenTh t position with respec constructioe th o t nucleaf no r power plants in the UK and prospects for the future Is summarised. completed The study has indicated that it would be technically feasible to build at least The paper reviews the effect that adopting high discharge burnup one nuclear power plant of up to 1500 MW(e) at Chapelcross If the project were to go cycles has on the fuel manufacturing process, the core design, the ahead, construction lanwouln o de db owne Britisy db h Nuclear Fuels adjacene th o t t reprocessine performancth n o fuee d th an ldischargef go f eo d fuel. existing Magnox station It Is concluded that there are no fundamental difficulties which would preclude high burnup cycles In the UK's PWEs, although a It is a safe assumption that any future programme of thermal nuclear power plant number of detailed design and licensing issues would need to be constructio mediue designsth R wil basen K i e mb lU PW e tern d,o th whicmn i h explains addressed. Some suggestion made sar e regardin contene e th gth f to UK'e th s interes Lighm t t Water Reactors (LWRs Goverment'e Th ) s decisio suspenno t d IAEA report on the impact of high burnup on the nuclear fuel cycle. the construction of the follow-on plants to Sizewell 'B' was a result of many factors The principal reason was due to difficulties in raising private capital to fund nuclear power plants, given the radical changes to the industry This in turn was cited to be due to 1. INTRODUCTION uncertainties over whether a secure market could be found for the output of the nuclear stations within the new electricity supply structure, but another important factor was cited The electricity supply industry in the UK has just undergone a major upheaval following to be the economics of nuclear generation Government'e th s implementatio plans it f removno so t e bot supple h th distnbutio d yan n sectors from state ownership Formerly Centrae th , l Electricity Generating Board (CEGB) Partly because of the restructuring of the industry, the cost of nuclear generation has was responsibl supple th distributiod r eyan fo Englann i Walesd dan , whil Soute eth f ho recently becom principae eth l factor whic figures publie hha th dcn i debate ove future rth e Scotland Electricity Board (SSEB) and the North Scotland Hydro Board were responsible of nuclear power in the UK The future of the UK nuclear industry will therefore depend to supple fo th distributiord yan Scotlann i d Following privatisation CEG splis intp Bu twa o n importana t exten beinn o t g abl demonstrato et economicalle b o t t ei y competitive three major suppliers, National Power, Powergen and Nuclear Electric and a separate Althoug nucleaa cose f th bule o tf hth o k r plant ove s lifetimit r s associateei d with company responsibl r distributioefo SSEe splis Th nB twa companiesinto otw , Scottish recoverin constructioe gth n costs fuee th ,l cycle costs mak significanea t contributioo nt Power and Scottish Nuclear, while the North Scotland Hydro Board was renamed the total (~ 10%) One means of improving the fuel cycle economics is to increase fuel Scottish Hydro Apart from Nuclear Electri d Scottisan c h Nuclear l thesal ,w ne e discharge burnups Another is to utilise Mixed Oxide (MOX) and recycled uranium fuels companies were privatise recene dTh t legislative changes have also mad et possibli r efo activels i K U y e pursuinTh l thesgal e options Higher discharge burnup thereforn sca e independent generating companies to be formed to provide competition to the mam contribute to the overall case for a continued nuclear programme in the UK generator alreadd san numbeya f privatelro y funded generating station beine sar g built This presentation review currene sth t position fuewitK R U hl cycl e PW regar th e en th i do t Such major changes to the electricity industry have naturally had a major impact on the identified an technicae sth l issues concerning high burnups whic consideree har d most UK nuclear industr e GovernmentTh y s original plans wer o privatist e e nucleath e r significan scopreviee e th Th f t fueewe o th coverl f cycleo impacfrone e d sth th ,en t n o t component f CEGso SSEd Ban B along witfossie hth l fuel components t thesbu , e plans e m-reactoth r fuel managemen d performancean t e impacth s weln a fues ,o a tl l were change lata t eda stagfinae th l, e decision bein separato g t nucleae th f eof r reprocessing components into separate companies, Nuclear Electric and Scottish Nuclear, and to retain them under government ownership Prior to the time at which the decision was mad o t retai ee nucleath n r generating stations under government controle th , 2 CURRENT STATUS OF PWR FUEL CYCLE IN UK Governments intentio construco t s nwa t four Pressurised Water Reactors (PWRsa f o ) Westmghouse design three in England and one in Wales by about the end of the Constructio f Sizewelno currentls i B l progresn yi scheduln o d r scompletioan efo y nb CD W centure samth t e A t i ydecide tim privatiso s t a et dnucleano e eth r generatione th , plane 199 WestmghousTh e bases i 4t th n do e SNUPPS design"' similas i e th d o an ,t r 03 Wolf Cree Callawad kan BritisA yUS planthe Nucleath n si r Fuel supplyine sar initiae gth l An issue which has not yet been addressed in the UK is the economic impact of utilising -t* charge fuel and have been responsible for producing the nuclear design report and MOX assemblie hign si h discharge burnup cycles Sinc manufacturine eth X g cosMO f o t carryinassociatee th t gou d safety related calculation firse th t r operatinsfo g cycle Th e assemblies is essentially independent of the initial fissile content, MOX fuel can initial core loading pattern is virtually identical with those used in the first cycles of Wolf potentially give economic savings when utilise higdn i h discharge burnup cycles sucs ha Creek and Callaway. with enrichments of 2 1, 2 6 and 3 1 w/o for the three fuel regions the above four batc mont2 h1 h fuel management scheme Since the first reload is not due until 1996, or thereabouts, the form of the reload fuel management has not yet been finalised by Nuclear Electric Nevertheless, comprehensive calculations of the performance and safety characteristics of nominal 4. IMPACT OF HIGH BURNUP FUEL CYCLES ON THE reload fuel cycles have been carriesatisfo t t safete dou th y y authorities thacore th t e FRONT-EN FUEE TH L F CYCLDO E characteristics in reload cycles will be acceptable 4.1 Impact on Fuel Fabrication These nominal reload studies, carrieBritisy b t hdou Nuclear Fuels, have been basen do a very conservativ month2 e1 , three batch fuel cycle ,achieve e whicb n hca d wit hfeea d Ther implicatione ear f higso h burnup fue fue laspect R cycleo ltw fabricatioPW n f so s o n enrichment of 31 w/o The discharge burnup depends on the load factor assumed and is in the region of 33 GWd/t Up to the present time, relatively little priority has been given An important consideration is that of the cnticahty clearance for the fabrication plant to establishing more advanced fuel cycles , the main priority has been to construct the British Nuclear Fuel's oxide fuel fabrication plan cleare s i tmanufacture th 5 r do fo t p u f eo plant to time and to cost and to obtain good operating performance This conservative w/o enrichment, a figure which is typical of modern oxide fuel fabrication facilities This approach is likely to be continued in the first reload cycles will be adequate to cope with fuel intended for burnups of up to at least 45 GWd/t, even In the longer term, however, it is likely that the priorités will change and that there will be allowing for split enrichment feeds With discharge burnups significantly higher than 45 more emphasis on more economic fuel cycles, in both Sizewell 'B' and later PWRs, as GWd/t, careful attentio cnticahte th o nt y issue would give e havrequiremena b f I o net t has already been considere r BNFL'dfo s proposed PWRs British Nuclear Fuels have arise r fuelsfo s enriched beyon dw/o5 application a , n would e made havb th o o et t already reviewed extensively high burnu fueR l pPW cycle havd san e concluded thaa t licensing authority This would involve at the very least re-calculation of the potential for four batch 12 month cycle would give substantial benefits in terms of fuel cycle costs'2', alsy criticalitoma requird yan e revised procedure implementede b o st addition i , o nt as discussed in the next section significant cost penaltie plann si t equipmen t presentt cleaA tno rs i whethe t i , r there ear any practical considerations which will limit the highest enrichment levels manageable and what level of enrichment and discharge burnup such a practical limit would 3. URANIUM FEED, ENRICHMENT AND OVERALL FUEL CYCLE COSTS OF correspond to. 4 BATCH 12 MONTH CYCLES f higo Thhe eburnuus p fuel assemblie alsy osma have implication manufacturine th r sfo g A four batc mont2 h1 h fuel cycle woul wele db l suiteannuae th o dt l load demand cycle, plan neea than i t e provido tdb t thery ema integral fuel burnable absorbers, sucs ha peae th whic ks demanhha wintee th n di r minimu e monthth d man s demane th n di Gd2O3 rods, in which case a separate manufactunng facility for the integral poison rods summer BNFL's study'21 has shown that a four batch 12 month fuel cycle in a 1100 may be necessary This will certainly increase the unit costs of fuel fabrication, but MW(e) plant wit dischargha e burnu f aboupo GWd/4 4 t t would reduc feee eth d uranium probabl suco t extent n hinvalidato a t y no s a t fuee eth l cycle cost benefit f higso h burnup and conversion requirements by about 3% relative to a 3 batch 12 month cycle. With the cycles feed enrichment increase w/o4 enrichmene 4 th , o t o d w/ fro tfoue 1 costmth 3 r r sfo batch cycle would, however, be some 4% higher, so that the combined uranium feed, Impac2 4. Fuen to l Management conversion and enrichment costs of the two fuel cycles would be about equal On the assumption thafabricatioe th t spend nan t fuel management cost r unispe t weigh f fueo t l The increased initial enrichments needed for high burnup fuel cycles have implications for are approximatel fueo tw l cycles e same th y th higr e efo th , h burnup fuel cycle shown sa fuee th l management Increasin initiae th g l enrichment introduce a larges r reactivity advantag approximatelf eo fuen i l % cycly13 e costs differential between fresh and previously irradiated fuel, which tends to cause the radial power peaking factor riso st e Dependin radiae th n lg o peakin g factor limits applicablo et There is thus a substantial benefit to be gained from increasing the equilibrium discharge a particular plant, and on the fuel management scheme (le whether out/in or low burnup of the UK PWRs to = 45 GWd/t Provided that the safety characteristics of high leakage), this may necessitate the use of burnable absorbers for radial flux/power burnup cycles prove to be satisfactory when evaluated against the UK s requirements, peaking control this will be a major incentive for the adoption of such a fuel management scheme, or a PWRK U e s th Preliminarn similai e on r y studie r highefo s r burnup fuel cycles confirm Sinc reloae eth d fuel managemen beet ye nt PWRK tno schemU s e usesha e th b dn o ei t that ther continuee ear d fuel cycle cost saving t evea s n higher discharge burnupt bu s decided it is too soon to say whether burnable poisons will be needed with high burnup recognisin neee gwelr th dfo l planned gradual steps higher discharge burnup cyclee sar cycle f thiI s cas se th prove ee b gadolmi o st a burnable poiso wely favouree nb ma l s a d not implementelikele b o yt PWRK U dn i s immediately it allows separate control ovee initiath r l reactivity hold-dow burn-oue th d nan t rate Table 1 the overall impact on safety case calculations roughly neutral Other reactivity coefficients, suc Dopples ha borod an r n coefficients affectee ar , impacde alsoth t n o tbu , Impact of High Burnup Fuel Cycles on Core Characteristics core th e managemen less i t s significant tha nmoderatoe thath f o t r coefficient The reduced control rod worths in high burnup cycles arises because of the increased Parameter Effec f Higo t h Burnup Cycle Implications thermal absorptio hige th h n ni enrichmen t fuel Combined witmore effece th hth ef o t negative moderator temperature coefficients the reduced control rod worths tend to Radial peaking Increases Reduced margin to reduce the shutdown margins This needs careful attention in the fuel management to factoH 4 rF FaH limits ensure tha shutdowe tth n margin t reduceno e sar unacceptabldo t e levels Axial peaking factor No significant effect A comprehensive assessmen impace th f fuen o t o t l managemen f higto h burnup cyclen si t beeye nt PWRcarriethno K eU s t Preliminarsha d ou y studie4 a r sfo reporte2 f Re dm Overall peaking Fo Increases Reduced margin to Fo batc mont2 h1 h cycle, however, have identified tharadiae th t l peaking factor aspecf o t limits fuel managemen moss i tlimitin e th t likele b g o yconsideratiot n Moderator temperature More negative Reduced shutdown margin Table 1 summarises the impact of high burnup fuel cycles on the principal safety related coefficient Increased feedback in core parameters, identifying against each paramete parametee th w ho r r changee th s sa heat-up faults Increased discharge bumup increases reactivity insertionn i cooldown faults Impac3 4. FueDesign d to Ro l n Dopplesignificano N borod ran nt effect For discharge QWd/t5 burnup4 o t ,p swhicu highese th hs i t burnup level which British coefficients Nuclear Fuels envisage to be likely in the forseeable future in the UK PWRs, no new fuel performance issue expectee sar fuee th arisl o d t performancd ean e safety assessments Control rod Reduced Reduced trip reactivity expectet no affectee e b ar o dt d provided thaexistine th t g core design limit satisfiee sar d worths worths Reduced shutdown At discharge burnups extendinGWd/t0 6 fuee o t on l,p gu performanc e issu bees eha n margins identified to be limiting, that of fuel cladding corrosion Xenon reactivity Reduced Improved operating The increased dwel fuele timhign th i l f eho burnup fuel cycles combination i , n wite hth worths flexibility higher rating whico t s fuee hth l wilsubjectee b l t higda h burnups, will increase eth potential for clad corrosion This was identified as an issue by a recent IAEA study on fuel performanc t higea h burnups'3' Severa fueR l vendorPW l currentle sar y developing improved cladding materials, such as modified zirconium alloys and duplex cladding (in whic Zircaloe hth y cla coates di d wit corrosioha n resistant layer) British Nuclear Fuels Gadolmia poison requiro sd e careful attentio ensuro nt e that within-assembly peaking are monitoring developments in this area closely, and may choose to adopt such an effecte th f d reduceso an d thermal conductivit t becomno o yd e limiting t thesbu , e effects advanced claddin cleaa gf i rshows i nee t i r exisdno t fo t can be minimised by careful selection of the UOa enrichment in the gadolinia rods With modern nuclear design codes, questions regarding the prediction of gadolinia behaviour Other secondary fuel performance issues, which arise from the more severe rating should not be a major consideration histories that fue subjectes i l hign i ho dt burnup cycles, rather than fro performance mth e of the fuel itself, are those of fission gas release and pellet clad interaction The higher enrichments used in high burnup fuel cycles also have a significant impact on the reactivity coefficients and control rod reactivity worths Increasing the enrichment Although British Nuclear Fuels' Integrate y RoutdDr e (IDR) fuel shows excellent fission tend mako t s moderatoe eth r temperature coefficients more negative With respeco t t product retention properties e longeth , r exposur ee fue conjunction th timi l f eo n with beginning of cycle moderator coefficients this is a beneficial effect as the requirement to increased rating t higsa h burnups increase potentiae sth r fissiofo l releass nga e British have a non-positive moderator coefficient at operating conditions can be met more Nuclear Fuels, in common with other fuel vendors, are assessing this potential problem easily, with fewer (if any) burnable poisons required for moderator coefficient control The and methods of reducing fission gas release more negative moderator temperature coefficients may have a detrimental effect on cooldow npotentiae th fault s a s l reactivity insertio s increasen i othe e th rn O handd , Finally, there is the issue of pellet clad interaction There is a clear need to demonstrate CD reactivity feedback in heat-up faults is increased, which is a benefit and tends to make that the failure threshold for this mechanism does not become significantly more CD restrictiv t higea h burnup thad combination si tan n with higher local rating hign si h burnup 6. RECOMMENDATIONS FOR THE IAEA HIGH BURNUP STUDY O) fuel cycles, doe t resulno s excessivn i t e number f fueso l failure postulaten i s d reactor faults This presentation has summarised the current status of high burnup LWR fuel cycle work in the UK Following from the points made in this report, the following recommendations are made regardin contente gth IAEe th Af so impace reporth n f higo o tt h burnupn so 5. IMPACT OF HIGH BURNUP FUEL CYCLES ON THE LWR fuel cycles BACK-END OF THE FUEL CYCLE IAEe Th 1A report should attemp determino t t e whethe r practicay theran s ei l limin o t s prematuri Idecido t t whethen ye t eo fue s PWRno K ea l U r fro o re smth wile b t enrichment due to criticality considerations in manufacturing plants If so, the report reprocessed Nevertheless, the option to reprocess will be available in the UK in British should identify approximatel t whaya t burnup level sucy san h limitation woul effective db e Nuclear Fuels Thermal Oxide Reprocessing Plant (THORP) THOR currentlPs i y being repore Th t should identif y speciayan l measure snecessare b whic y hma preveno t y t license r operatio dfo GWd/0 4 o t t n p dischargfuel R u wit f so hLW e burnup Following criticalit n manufacturini y g plants oved abovan r e those already require t presenda t from a preliminary assessment of the safety implications of reprocessing high burnup fuel burnup levels (up to 60 GWd/t) in THORP, it has been concluded that with suitable changes in operating procedures, a satisfactory safety case can be made The following discussion IAEe Th A 2 report should includ surveea effece th f higf o tyo h burnup e cycleth n so highlights the most important areas in the reprocessing plant which are affected by high core neutronics characteristic identifd s an factor e yth s whic mose har t limitinlikele b o yt g burnup fuels from the perspective of fuel management and operational flexibility An assessment of the fuelX hign economii sMO hf o discharg e cus impace th e f burnuo t p cycles woule db Mechanical handlin d feean gd pond operation e mechanicath t a s e lar head en d valuable The report should include a concise discussion of the relative merits of short (•= unaffectedprocessine th e ar s ga , rate cycld san e mechanica e timeth r sfo l operationsn i month4 2 1 2o t month)utilito 8 t fue1 lond d y= lan ai g( ) manager cycles n a s a , n si the basket handling cave Dose uptakes, shielding, cooling and criticality considerations selecting fuel management schemes in these areas are, however, affected and would need to be assessed IAEe Th A 3 report should incorporat ereviea fuee th l f wperformanco e implicationf so Dissolution times in the chemical head end plant are unaffected, though there is an high burnup cycles based on the conclusions of the IAEA study on fuel performance at increase in the amount of insoluble fission products in high burnup fuels, which will affect high burnups reported in Ref 3 the amount of feed liquor volume processed per batch in the centrifuges This will not significantly affec on-line th t e cycle time, however, becaus off-line eth e tim r removaefo l of insolubles is relatively short compared to the on-line process time Preliminary studies REFERENCES indicate tha satisfactora t y safet ychemicae casth r efo l hea pland den t coul made db e with high burnup fuels 1 B V George & J A Board The Sizewell 'B' Design Nucl Energy June. 3 ,o 1987N , , 133-1426 , 8 Constraints related to maximum Pu capacities would lead to a reduction in trie maximum throughpu chemicae th n i t l separatio highee th nconten u o plantt P r e f higdu ,o t h burnup C Robbms. S M Connolly A High Burnu UtilisinR p CyclPW ga en i fuels A further consideration is a constraint related to maximum activity level of feed & K W Hesketh Gadolmia Burnable Poison liquor These consideration maky sma t ei necessar perforo yt m blending operationo st Proceedings IAEA Technical Committee Meetinn go ensure that the liquor derived from the high burnup material is diluted In-Core Fuel Management, Vienna, 4-7 Dec 1989, IAEA-TECDOC-567 Overall, the processing of high burnup fuel will give rise to no new effluent streams, and the ansmgs of solid, liquid and gaseous effluents will be similar to those of medium Fuel Performanc Higt ea h Burnu Water pfo r burnup fuels althoug volumee hth concentrationd san diffey sma r somewhat Reactors International Working Grou Waten po r Reactor Fuel Performanc Technologd ean y Thus therreasoo n higy es i nhwh burnup fuels shoulreprocessee b t dno THORdn i P with Studsvi Jun8 5- ke 1990, IWGFPT/36 suitable modifications to operating procedures BNFL is currently preparing a full safety justificatio r reprocessinnfo g high burnup fuel THORn additionasi o n d Pan l inpu thin i t s area is seen to be necessary at present EFFECTS OF EXTENDED BURNUP utilities extend burnu constant a p t cycle length while most U.S. utilities extend both burnu d cyclan p e length witexpectatioe th h ne longe th tha n i tr ON THE NUCLEAR FUEL CYCLE term (for "equilibrium cycles") reload fractions wilt changno l es i much t I . important to recognize that some of the effects of extending burnups depend on P.M. LANG which of these approaches is used. When such differences occur, the conditions United States Departmen f Energyo t , to which effects apply need to be clearly stated. Washington, D.C., . 2 Front-End Effects United States of America e front-enTh d effect dividee b n sca d into partstw o : effect n front-enso d resource requirements and the effects of the higher enrichments needed for Abstract higher burnups on fuel fabrication plants and fresh fuel storage facilities. Burnup affects to varying degrees the front end of the fuel cycle, the 1 Effec2. n Resourcto e Utilization ln-reactor part of the fuel cycle, and Its back end. These effects are reviewed and discussed in this paper from the US perspective of PWRs and In Reference (2), three measures of resource utilization were discussed, each BWRs operating with storage and disposal of spent fuel. The front end of which is defined in terms of the amount of energy produced per unit of the effects include reduced requirement uraniur sfo fued man l fabricatiod nan resource consumed. Historically, the most important and most widely used licensin front-ene th f go d operatio highee th r r nfo enrichment s needed measur s beeeha n uranium utilization. Other significant measure f fueo s l for higher burnups. The principal in-reactor issues are (1) the fuel utilizatio separative th e ar n e work unit (SkU) utilizatio fuee th l d an n desig performancd nan e impactlongee th f rso in-core residence timen so assembly utilization utilizatioU SW e Th . n characterize amoune sth f to corrosion and hydriding, fission gas release, dimensional and structural separative work neede unir f energpe do t y produced e fueTh .l assembly changes, pellet-cladd interaction, and ultimately the resulting effect on Utilizatio measura s i n botf eo h fuel fabrication and, more importantly, spent fuel reliability, and (2) the effects on fuel management of greater fuel generatio unir npe f energto y produced givea r nFo .fue l design alst ,i o potential differences in reactivity between adjacent assemblies and the imeasur a samoun e th metaf f to eo l (e.g. Zircaloy )fuee needeth l r dfo potential for reduced control-rod worth and changes in some reactivity fabrication process. Because of difficulties and high costs associated with coefficients. The principal effects on the back end are the reduced e back-en th e fueth l f o cycld mann i e y countries e fue,th l assembly utilization volume of spent fuel to be handled and the higher decay heat generation as a measure of spent fuel generation has gained strongly in importance and of this spent fuel. Finally, there is a need to be able to take credit emphasis in recent years. A fourth utilization may be similarly defined for for burnu licensinr pfo g spent fuel storag shippingd ean . conversion; sinc uraniul eal convertes i m hexafluorido t d e prio o enrichmentt r , e conversioth n utilization varie exactln i s e samyth e mannee uraniuth s ra m utilization. 1. Introduction 2.1.1 Effect on Uranium Utilization discussioe Th effecte extensioe th th f f o nso burnuf no lighf o p t water reactor e effecTh f extendinto g specifie burnuth n o p c uranium consumptio shows i nr fo n fuel on the nuclear fuel cycle presented here is divided into three parts: several reactor/cycle-length combination Fign i s . 3.4.7.1 WREBUe th f .o S cyclee th effectf o ,frone th d in-reacto ten n so r effects effectd e ,an th n so study (1). The specific uranium consumption is expressed in grams of natural cyclee th e bac discussiod Th f .o en k bacd en k n appliee once-througth o t s h uranium used per HW-hour of electric energy generated and may be used as an fuemore th le r cycleeconomifa y s thib ,a s i cse optio optioth d nan nwhic s i h indicato uraniuf o r m utilization e resultTh . Figf so . 3.4.7.1. show that being selected in most countries employing water reactors. A large part of specific uranium consumption goes down (i.e., that uranium utilization this discussio takes i n n from relevan tWREBUe partth f So s report being improves s burnu)a p increases, almost without exception. published by the IAEA (1). These sections of WREBUS were originally written by the present author; hence o usefu,n l purpose woul servee b d y attemptinb d o gt Comparison of 12- and 18-month cycles for the same reactor at the same burnup documenw writ ne e sam a th ee n tsubjectso incorporatinn I . g these sections show se longe th tha n i tr cycles, uranium utilizatio degrades i n y typicalldb y from WREBUS, editing, summarizing, and supplementing with additional 10-15 percent. A similar comparison of 18- and 24-month cycles for the U.S. information has been done, as necessary to adapt these sections to the BWR shows that this further lengthenin cycle th ef o gadversel y affects uranium organization and purposes of this paper. utilizatio percent9 betweey b nd an n6 . Burnup may be extended in different ways, as follows: (1) at constant cycle e lesTh s uranium-efficient performanc f longeeo r cycles arises froe pooreth m r length with accompanying reduction in the fraction of the core reloaded, neutron economy inheren e desigth n f longei o nt r cycles, which results froe th m ) whil(Z e extending cycle length proportiona o burnut l d keepinan p e th g need to load more excess reactivity at beginning of cycle and hold it down by fractio core th e f reloadeo n d unchanged ) whil(3 r e o ,simultaneousl y changing control absorbers. Burnup extension typicall e potentiath s o ha yt o sav t lp u e •-00J cycle length and reload fraction. As a first approximation, most European about IS percent in specific uranium consumption when implemented with 00 unchanged cycle length. For reactors operating at the historical burnups in Fresh fuel storage has been studied for PWR fuel of up to five percent 00 the neighborhood of 30 MWd/kgll, approximately half of this uranium saving can enrichment, and the results show that no criticality concerns exists for be realized if burnup and cycle length are simultaneously increased, to the typical storage rack designs. higher value f burnuo s 8 months p1 e figur showo th t d fron .i 2 nan e1 m 2.1.2 UtilizatioU EffecSW n to n 3. In-Reactor Effects of Extended Burnup e effecTh f increasinto g consumptioe burnuth n o p f SWUo r unin f energpe so t y generated is generally small, and the direction of the change can be either 3.1 Fuel Performance Considerations increasing or decreasing. In most studies which cover a wide enough range The principal fuel design and performance technical issues in extending burnup (3), the specific SWU consumption at first decreases, reaches a minimum at are fission-gas release and internal fuel-rod pressure, cladding corrosion and intermediate level f burnuo s p extension d thean ,n increases agai s burnua n s i p hydriding, fuel dimensiona d structuraan l l changes (rod growth, assembly extended to the highest values. The greatest difference found between an growth, rod bowing, clad creepdown or expansion, grid-spring relaxation, etc.), extreme value and the minimum value is only a few percent. Because of the and fuel integrity; i.e., resistance to pellet-clad interaction (PCI) and other relatively small magnitud mosf o ef theso t e changes e lac ,th f unifor o k d an m failure mechanisms. Host of these issues can be resolved through consistent trend thein i s r directio d magnitudean ne relativelth d an , y lesser straightforward design changes such as providing more space in the assembly for interest in SWU utilization generally, it is concluded that the effect of fuel-rod growth, morr fission-gafo e e fued spacth ro l n i e s releaser ,o extending U burnuutilizatioSW n po relativels i n y smal d unimportantan l . lowering the prepressurization pressure. Of potentially greater concern is waterside corrosion ;mose thith ts i slikel y fuel life limiting phenomenor fo n 2.1.3 Effect on Fuel Assembly Utilization present zircaloy cladding metallurgy r hig,fo h enough burnup extension and/or high enough coolant temperature. For a fixed fuel assembly design containing a fixed weight of enriched uranium, burnup itsel direca s i f t measur f fueeo l assembly utilizationt i s a , Advanced zirconium-based alloys with better corrosion resistance orden ,i o rt represent energe sth y produce r uni pe f denricheo t d uranium loadedo T . facilitate burnups higher than achievable with present alloys, are now becoming calculate a specific fuel assembly consumption analogous to the specific commercially available. Fuel failure through PCI is not a significant concern uraniuU consumptioSW d an m n discusse e precedinth n i d g sections, burnup would at extended burnup, both because advanced designs that enhance PCI margins are need to be multiplied by the weight of enriched uranium per assembly and by the available and because at higher burnups the local reactivity is likely to be thermal efficiency of the plant, in order to obtain electrical energy per insufficient to produce power ramps that result in risk of PCI. The PCI assembly reciprocae th d ,an f thio l s product would represent assemblier pe s concer mor s idli l n eva y directed towar e performancth d t loweea r burnupf so unit of electric energy. This calculation is hardly worth doing, since burnup higher enrichment fuel designe achievo t d e extended burnup; this performance e onlith s y variabl d thereforan e direca s i e t measur f fueo e l assembly must still be maintained free of PCI failures through PCI-margin enhancing utilization. design improvements, lower linear heat rates, or control of fuel local power and power increases. The specific assembly consumption is proportional to the reciprocal of the burnu thereford an p e always decrease burnus a s p increases r smalFo o .t l A traditional concern when extending burnups has been the potential effect on moderate changes givea , n percentage increas burnun i e p gives virtualle th y fuel reliability. There exist n instinctiva s e feeling that operatin e fueth gl same percentage increase in fuel assembly utilization. This explains the longe imposins i r mora g e sever whic, e it dut hn o yshoul d lea o increaset d d significance of burnup extension in reducing the amount of spent fuel which fuel failures .factn i y experienceb t bornThi , ,no t is sou e . Reliabilitf o y musreducinn i e dispose b td an gf o dassembly-specifi c costs, both fuel fuel (an f mando y other products bese b ty )understooma consideriny b d e th g fabricatio back-endd an n . traditional bathtub-shaped curve of failures vs. usage or time in service characterizin occurrence th g f failure eo componeny an f o s r systemto e Th . 2.2 Effects of Higher Enrichments initial, sharply decreasing part of this curve represents the preponderance of manufacturing and quality assurance related defects early in life. This is For some fuel suppliers, processing higher enrichments has required or may followed in the intermediate range by a low, horizontal line representing the require amendment of their fuel fabrication license. While this may require effec a smal f o tl numbe f randoo r ofted an m n external cause f failureo s . time and documentation, no difficulties are foreseen for the moderate increases Finally, late in life, wear-out effects predominate and the failure curve in enrichment presently contemplated. The current license limits on enrichment begins to rise sharply. With regard to nuclear fuel, its technology must be for the five U.S. fuel fabricators are: well understoo s desigit d n an dconservativ e b d well-tested o an et s i t i f i , license a substantia n o d l scale. This means thaoperatios it t always i n s ABB-Combustion Engineering 5.0% restricted to well below the point at which wear-out effects begin to take Babcoc Wilco& k x 4.1%, being raised to 5.1% place. Hence, reliability has not in practice been found to decrease with General Electric 6.0% increasing burnup. In fact, for fuel with high rates of initial failures, Siemens Nuclear Power 5.0% reliabilit y welma y l improv s burnua e s increasedi p e effectth ,f o sincs e eon Westinghouse 5.0% f burnuo p increas a reductio s i w fuee amoun ne th l n f i o loadetn d each year. To establish utility confidence and regulatory acceptability for extended f cycleo d becaussan e reducing end-of-cycle absorber residuale b y ma s burnup fuel desig performanced an n , test assembly irradiation e desireth o t sd important. Low leakage fuel management becomes increasingly important as maximum assembly burnups contemplated are helpful prior to full implementation burnup (and reload fuel enrichment e increased)ar , since placing higher of extended burnups by utilities. This includes taking existing designs to enrichment reloa periphere d th fuee corunacceptabln t th a l eca f o y y increase higher burnup d testinan s g new, advanced designs throug procese th h f designo s , neutron leakag producd an e e excessive fast neutron fluencreactoe th n o e r fabrication, test-assembly licensing, irradiatio higo t n h levels d postan , - vessel. Another fuel management concer transitioe e desigth th s f i no n n from irradiation examination. Many such projects have been successfully lower burnups and out-in fuel management to extended burnup with low leakage accomplished for the extended burnup levels now considered acceptable (about 50 fuel management, either wite samth he numbe f corro e region cyclef i s e sar MWd/kgU for PWRs, 45 MWd/kgU for BWRs), but new projects would facilitate simultaneously lengthened r wito , h increased numbe f regiono r cyclf i s e length extension of burnups beyond these levels. is unchanged. Such transitions can be difficult to design if it is desired to achievw conditionne e th e s rapidly. Alon e samth ge lines e severa, th dat n o al aspect f fueso l performanct a e burnups higher than the established range of present analytical methods and 3.3 Licensing Aspects computer codetestee b y nee o sma t dd against these method codesd an s . After any necessary modifications, this will establish the validity of the methods A number of safety, licensing, and environmental assessments of burnup over the newly extended burnup range of interest. extension have been conducted in the United States, and as a result of these efforts fuel with progressively increasing levels of burnup has been licensed Burnup extensio y requirma n e some dimensiona materialr o l s changee fueth l o t s ove lase yearsw rth e fe overalt Th . l conclusion varioue th f so s assessments designs, and such changes, as well as the costs of extending warranties to are: higher burnups, recovering applicable research/development/testing/engineering o qualitativelN - y different phenomen e expectear a t highea d r burnups; costs, and in some cases increases in the amortization of capital costs over - No substantial safety issues exist; smaller volumes of fuel fabrication may be expected to increase unit - The environmental effects of burnup extension are mildly positive; and fabrication costs. Estimate thesf so e cost increases have beee nth mad d ean - Substantial licensing-oriented work may be needed whenever burnup is conclusion seems to be that such increases have been and are expected to be extende o newdt , higher levels—to obtain dat o supporat t licensino t d an g quite moderate moro ,n e than about 10-15 percent. show that these data either justify the extended burnup fuel operating within previously established licensing limits, or to provide technical 2 Fue3. l Management Issues justificatio y requirean r fo nd extensio f theso n e limits. Extending burnup requires increasing the enrichment level of fresh fuel and Most of this previous work has applied to burnup extension up to 50 MWd/kg results in lower enrichments and reactivities of discharged fuel. Physical discharge batch averag MWd/kg5 4 r PWR d efo r BWRs an s fo Uthero t n ,bu s i e consequences of this include: reason to believe that the principal conclusions would be substantially differen r furthertfo , moderate extensio burnupf o n , although coursef ,o , this - larger potential differences in reactivity between neighboring assemblies remains to be proven. e coreith n , resultin greaten i g r difficult maintaininn i y g power distributions within prescribed limits, and 4. Back-End Effects - hardening of the neutron spectrum, which reduces control-rod worth and The most important effect of extending burnup on spent fuel storage, shipment shutdown margin and may adversely affect some reactivity coefficients. and disposal, arises from the lesser quantities of spent fuel generated at higher burnups s indicateA . Section i d n 2.1.3 r moderat,fo e percentage Because of these changes, the constraints in working out acceptable core increases in burnup, the amount of spent fuel is reduced by the same management pattern r highefo s r level f burnuo s p become more severe; thiy ma s percentage. Thus, there is less spent fuel to be stored, shipped, and disposed also affect some safety analyse cord an se maneuverability. Experienced fuel r highefo f ro burnups. Existing storage capacities last longer before becoming management personnel have, however, generally been able to develop acceptable full, fewer spent fuel shipments need to be made, and any given repository fuel loading patterns for substantial increases in burnup. tonnage capacity will accommodate spent fuel whic s produceha h d proportionately more energy e incrementaTh . r variablo l e cost f theso s e operatione ar s Burnable absorbers are not needed for burnup extension by itself. When longer approximately proportional to the amount of spent fuel; hence, burnup extension cycles and/o w leakaglo r e fuel managemen e use conjunction ar ti d n with burnup yields substantial savinge varioue costth th f n so i ss back-end operations. extension, burnable absorbers become necessar r acceptablfo y e fuel cycle designs; specifically o hol,t d down excess reactivit beginnint a y f cycleo g o ,t Spent fue f higheo l r burnup highes ha s r decay heat generation rates than spent control power distributions more precisely, to reduce end-of-cycle absorber fuel of lower burnups, when compared at equal cooling times after discharge residuals o maintait d an , n reactivity coefficients within desired rangese Th . from the core. This is of importance because some of the proposed repository use of improved integral burnable absorbers may be particularly important for designs are heat-load limited. Spent fuel shipment and disposal in the U.S. is CD substantial cycle length extension of PHRs, as there may be no acceptable expecte o takt d e placa comple n o e x schedule reflectin e relativth g e CO alternative way o controt s e largth l e excess reactivity e beginninneedeth t a d g availabilit f speno y t fuel storage spac varioun i e s utilities' spent fuel CD poolsa firs s A t. approximation, this wiîl probably meana thar fo t No major technical problems or issues are known which would prevent moderate O considerable time after shipments from plant sites begin, these shipments will extension of burnup for the majority of existing plants. For some PWRs e madb e only from plants with virtually full pools y extendinB . g burnupa , operatin e highesth t a gt coolant outlet temperature usind an s g zircaloy utility decrease annuas it s l rat f dischargo e f speno e t fued thereban l y cladding, corrosion may limit the permissible degree of burnup extension. extend perios it s d before pool fill, proportionatel extensios it o t y f o n burnup morA . e vali practicad an d l basi comparisor sfo decaf no y heat No substantial safety or environmental issues are known to exist for burnup generatio t differena n t burnup r spenfo s t fuel shippin d disposaan g s i l extension, although substantial work may be needed to support licensing to therefore at cooling times proportional to burnup. When compared in this way, higher burnups. The net environmental effects of burnup extension are mildly spent fuel at extended burnup still exhibits higher decay heat rates, but by positive. much less than when compare e samth e t coolina d g time. Several alternatives r takinfo g car f thio e s remaining extra heat generatio thosn (i n e cases where The major benefits of burnup extension arise from the reduced number of fuel y actuallima t limitinge b y ) exist; these range from blending wite h th som f o e assemblies needed to generate a given amount of energy. This reduces the d colol an de sitd th o simplfue t et a l y providing longer cooline g th time r fo s amoun fabricatee b f fue o tamoune o t lth d f speno tan d te stored b fue o t l , highest burnup assemblies. shipped d disposaan , . of l In additio theso t n e effect f quantitso spenf o y decas tit fueyd heaan l t generation rate, spent fuel storag d shippinean n presenca g t criticality References concerns under the overconservative licensing assumption that all fuel be considered fresh; i.e., without credi r reducefo t d reactivity resulting from burnup. Such takee credib measurementf y i nt ma s verifyin burnue th ge ar p Report on Water Reactor Extended Burnup Study (WREBUS), IAEA, Vienna—in made, or it sufficiently stringent administrative controls are provided to publication. satisf e licensinth y g authorities that credi r burnutfo p shoul allowede b d n I . the United States, a burnup meter has been developed and successfully testd for P.M. Lang, "A Forward Look at Improvements in Fuel Utilization and verifying burnu f speno p t fue lburnua assembliesf o pe meteus e rTh . wit h Performance", Nuclear Power and Safety, Vol. 5, p. 491, International spent fuel of higher initial enrichment should resolve the criticality issue Atomic Energy Agency, Vienna, 1988. r spenfo t fuel storage neededf (andi , r spen,fo t fuel shipment. r spenFo ) t fuel shipment, the limiting factors to be considered for each fuel design and J.J. Maur t al.e o, "The Environmental Consequence Highef so r Fuel Burnup, shipping cask combinatio e heaar n t rejection capability, shielding, fission-gas "AIF/NESP-03Z, p. B-6, Atomic Industrial Forum, Washington, O.C., release criticalityd ,an e resultvere Th b . yn compleca s x sinc mano s e y June 1985. combinations exis d sinc an te sam th e e consideratio t alwayno s i ns limitingt A . the ris oversimplificationf o k statee b y dma that r U.Si , .fo t cask designs almost all fuel assemblies to burnups of 55 MWd/kgU (PWR) and 50 MWd/kgU (BWR) can be shipped in almost all of the existing casks, if one or more of the following are acceptable in some instances: 1) credit for burnup is taken in the criticality analyses, 2) somewhat longer cooling times are acceptable prior to shipping, 3) partial cask loadings are used in lieu of longer cooling times when these are unacceptable, and 4) minor design modifications are made t ocaskse th som f .eo Some relicensin highee th caske r th r fo sf burnu o g p fuel may be needed. For burnups beyond those cited, it should be expected that more of these measures must be taken if existing casks are used. However, the very high burnup assemblies under consideration for the future will not actually be discharged from reactors and need to be shipped for many years, and in that time fram ew desigcaskne f o sn anticipatin muce th gh higher burnup requirements should become available, obviating the need for excessive cooling times or partial cask loadings. 5. Conclusions Burnup extension reduces uranium consumption—but more so at lower burnups than at the highest values. Although cycle-length extension alone increases uranium consumption, simultaneous extension of burnup and cycle length can achieve about half of the uranium saving possible through equivalent burnup extension at constant cycle length. Other front-end effects are similarly favorable or neutral. IMPROVEMENT OF WER FUEL BURNUP the activities along these lines will be coordinated, as before, by AN EFFECFUES E DIT TH L N CYCLTO E the Ministr f yo Nuclea r Powe d Industran r supervised an ye th y b d USSR State Committee on Supervision of safe operation in industry Yu.K. BIBILASHVILI*. G.L. LUNIN**, and nuclear power. V.D. ONUFRIEV*, V.N. PROSELKOV** 'All-Union Scientific Research Institute of Inorganic Materials, 2. EXPERIENCE IN OPERATING WER CORES WITH ENHANCED Moscow FUEL BURNUP "I.V. Kurchatov Institute of Atomic Energy, Moscow 2.1. WER-440 Unio Sovief no t Socialist Republics Early WER-440 loadings were intende -3-yeaa r dfo r fuel lifetime at an average enrichment of 3.25% with out-in fuel transfer scheme. Abstract e refuelling-averageTh d fuel burnu 29MW.kg/daf po achieves ywa d with the cycl 2 eff.daye29 s <7000hr> long. The paper discusses the effects of burnup improvement on the WER The experiments on switching WER-440 to a 4-year lifetime of fuel cycle. The experience gained in operating the cores of WER-440 3.6%-enriched fuel performed in early eighties at Kola NPP proved to and VVER-1000 reactors wite focuth h n ssafetyo , reliabilityd ,an be a success as well as subsequent experiments with 4-year fuel economy issues raised in transferring to an enhanced burnup. The operation at Units 3 and 4 of Novo-Voronezh NPP and Units 1 and 2 of way optimizo t s e VVER cores, fuel assemblies fued ,an l elements with KolwherP aNP e fue alss lwa o enriche 3.6o t d % C2]. the aim to improve the fuel burnup and to improve the reactor safety are analyzed. In the paper, the secondary are the effect» of The increase of fuel enrichment up to 4.4% haa allowed not only enhancemen finaa n to l stag nucleaf eo r fuel cycle. the increase in the fuel burnup but also the increase in the cycle duratio .-**0 eff.dayo t 35 p u n -OÖOO s( wit> O hr h4-yea a r lifetime8 .7 fuel assemblies with 4.4% - enriched fuel (1/4 fraction of the whole I. SOVIET NUCLEAR POWER IN 1987-1991 core) were for the first time loaded to core periphery of Unit 3 of At present the NP fraction in electric energy production in the Kola NPP. USSR is about 12%. 47 NPP units of total installed capacity of 36412 In 1990 a 4-year lifetime corth ef o wite h enhanced-enrichment MW(e) were under operation in this country as of January 1. 1991. fuecompleteds wa l . Operating assemblies contained 4.4% -enriched structurthP N e e bein followss gl6000MW(ea ( R unit6 :1 WE 8 f . so )) fuel, while fuel sections of control and protection assemblies unit WER-44f o s 0 (3412MW(e)> 4 unit,1 RBMK-100f so 14000MW(e( 0 . )) contained 3.6%-enriched fuel- The duration of fuel cycles were 246. anunit2 dRBMK-150f o s 0 (3000MW(e)> e capacitTh . y factor (CF) 336 ,0 eff.days 37 300 d e averag,an Th . d maximuan e m burnuf o p averaged over this perio s 6t'-70%dwa . unloaded fuel assemblie s 42Mwa sW day/k d 48Man g W day/kg, e ChernobyTh l accident gav ewid. g birt eo t oppositioh P N o t n respectively. developmen e USSRth n .i tThi s P sloweNP w d ne dowe rat th f no e The experience gained served the basis for switching WER-440 commissioning resulta s A . prograe Soviee ,th th developmenP n tN o m t reactors to a 4-year fuel cycle involving 4.4%-enriched fuel. In s ha been revised A A drafanev. f o )t Nuclear Power Program orde validato t r future th e e transferrin 5-yeao gt r fuel cycle2 ,1 concentrate furthee th n o sr improvemen safetf o t performancd yan f eo fuel assemblies of Kola NPP Unit 3 remained under operation for the operating NPPs and on the development of improved-safety VVER-1000 fifth year of service. This experiment validated by preliminary units incorporating passive safety systems [13. It is supposed that thermal-physics and neutronics calculations proved to be a success. (O 22 WER-100 0 fue consumptio e lday/kW th M burnud 2 e improve4 b gan o o pt t f no p du N> The early VVER-1000 loads were intended for a 2-year tuel cycle natural uranium to be reduced by <*>1Z% [43. averagn a t e tuel burnu 29Mf po W day/k enrichmend gan 3.6%f to t .A The switching of VVER-1000 reactors to a 3-year, 4 4%-enriched present VVER-1000 reactors are being switched to 3-year fuel cycle fuel lifetime allowe e fueth ld burnu e improveb 43-45Mo o t pt p du W at an average burnup of more than 40MW day/kg and average enrichment day/kg and the consumption of natural uranium to be reduced by **j 15% of loaded fuel of 4 23%. [43. Novo-Voronezh NPP Unit 5 and Kalinin NPP Unit 1 have already abovementioneo tw e Th d switching have already been accomplished. reache steady-stata d e condition With three refuellings durine gth Below presented are some predictions [4J- core life e burnuth ; p averaged ovee fuellinron s reacheha g e dth - the reduction of deleterious neutron absorption in WER cores valu 43-45Mf eo W day/kg. by replacing steel in assembly parts with zirconium and the e implementatioTh 3-yeaa f no r fuel cycl Zaporozhyt ea UniP eNP t reductio hafniuf o n m conten zirconiun ti m will allo improvemenn wa t 5 was started with the first fuelling. The other operating VVER-1000 of burnu y b 2-3p d 8-10an X r XVVER-44fo d WER-1000an 0 , unit undee ar sr transferrin go 3-yeat fro- 2 rm core lifetime. respectively, or the reduction of loaded fuel enrichment by »\*0.4% Detailed information concerning this problem is given in [33. with respec o U-235t . Thiy provid ma sreductioe th e naturan i n l In order to validate 3-year lifetime, fuel element and assembly uranium consumptio y 3-7nb d X8-12an r WER-44Xfo WER-1000d 0an , respectively; operational reliability was both predicted and measured in experiments and their parameters were optimized. This will be implementatioe th - refuellingt o n s with reduced radial neutron considered below t presenA - e wholtth e comple calculationf xo d san leakage provides the additional improvement of fuel burnup by 5-6% experiment n progresi s i s validato st 4-yeaea r fuel lifVVERr to e - and similar reductio naturaf no l uranium consumption; 1000 reactors. the application of integrated burnable neutron absorbers in During scheduled preventive maintenance essentially each VVER- WER-1000 fuel load will allow the core arrangement with reduced 1000 uni s testetwa r fuedfo l clad integrity. Ove perioa r d from radial neutron leakag thin I es case fuee ,th l burnu increasy pma eb 1982 to 1990 4190 assemblies were tested and only 2 assemblies were y <•* 6-9X [5] and the natural uranium consumption mav reduce bv a unloaded befor e scheduleeth d a timesinglt No .e assembly failed similar value after three years of operation. Calculation R fueWE n lo s cycle optimizatio termn tixea ni f so d cycle duration (12 month8 ,1 WER-44r month2 fo s1 VVERr d sto Üan - 3. EFFECT OF IMPROVED BURNUP ON DEMAND FOR NATURAL 1000) were performed withi framewore nth IAEe th A f REBUo k S program in 1989-1991 URANIU ENRICHMEND MAN T LEVEL The dependence ol a tuel enrichment required for achieving a The experience gained and estimation performed show that in an particular tuel burnup (Fig.l s determine)wa d basin n physico g s open tuel cvcle (without recyclin t o reprocesseg d uraniud an m calculations of core refuellings with out-in fuel transport Physics plutonium e optimath ) l leve f fueo l l burnu n existini p R WE g calculations were experimentally verifie o fuet lp d u burnu f p45Mo W reactors is about 50 MW day/kg, in contrast to earlv design burnupa day/kg change WER-100a Th n .i e o 0t curve du es i characte) 1 g Fi ( r of 29-32 MW day/kg The present-day values will help to reduce the a specific core arrangement. Witincreasine th h g fuel burnud an p demand for natural uranium by about 20% [4]. associated increase in fuel enrichment, a core with a partly reduced Thus. the switching of VVER-440 reactors to a 4-year tuel radial neutron leakage placemene getth so t formef partlto e du dy lifetime where enrichment of loaded tuel was 4 4% in operating fuel outburned assemblie o st particula r position core th e n i periphers y assemblies and 3.6% in control and protection assemblies allowed the even if in general the tuel transportation scheme is out-in 34 0 7 5 6 0 6 5 5 0 5 5 4 0 4 5 3 0 3 5 "2 Burnup, MW dayAg 1 Enrichmen g Fi s fuev t l turnu5 16 p WEFM40— —• month2 1 , s —\— WHM40. 18 months WïR-*MO month8 1 , s — * WEK-1000— month2 ,1 s 26 L In this case the core optimization is performed to achieve a , uniform power density rather than the beat fuel utilization {61. 12 Thes approacheo tw e s differ most essentiall burnue th n pi y rangt eo up to"40MW day/kg. In the REBUS program, the fuel burnup effect on the consumption s predicte wa unite th enrichmenf o s n i d t work (SWU VVER-44r Fo ) U the value of S^U/KW,hr falls with the increasing burnup to medium values followed with a slight rise, the maximum difference being not more than 7% f7] For WER-lOOO the value of &WÜ/KW hr. reduces 20 WÏR-1000. 12 months monotonously with the increasing burnup. the difference achieves 21 556 when the burnup increases from 33 to ~65MWday/kg [7] The data on the VVER-440 and WER-lOOO fuel burnup effect on 18 natural uranium consumption estimated withi framewore e th nth f o k REBU e reductioSe Th naturath progra 2 n i g nls showFi i m n ni uranium consumption estimated by REBUS is rather close to earlier data predicted by Soviet specialists tor particular cases ot WER 16 switching from 3 - to 4-year fuel lifetime 111 5 and 12%, 0 7 0 6 0 5 0 4 0 3 respectively WER-lOOd an ) O switchin o 3-yeat g - fro r2 m fuel 80 90 100 Burnup, MW dayAg lifetime ( ~ 14 and -^1555, respectively) WER fuel burnup effect on tuel cycle economy xs not considered at the present meeting Detailed information on this issue mav be Fig 2 Specific natural uranium consumption CO vs fuel burnup for WES [7] W 3 foun[7 n i d CD 4. FUEL HANDLING IN THE CORE AND MONITORING OF WER REACTOR R reactorsWE In value sigd ,th coefficienef an no reactivitf to y J REACTIVITY BEHAVIOUR IN THE PROCESS OF FUEL BURNUP INCREASING in coolant temperatur OkT( e =depen> &jff)T critican do * ~ C l concentration of boric acid in coolant, startup position of a 4 1 WER-440 control group, core load composition, degree of assembly burnup. and on a very temperature Core characteristic structural materials are In the beginning o± a tuel cycle the VVER-440 and WER-IOOO core such thaVVER-44r tfo 0 load ratef so d energy capacity (29 davsf 2ef . possesses a maximum reactivity margin intended tor bringing the combinatioy an t a f ) startuo n6% fue3 pld an enrichmen % 4 2 f to reactor to an operating condition. Fuel burnup during 7000 hr (292 parameter e beginninth t fuea a s lf g o negativs cycli j of e e oven ra eff.days a nomina t a )l powe s 137i r 5 MW(th d SOOOMWrth)an ) , operating temperature rang o 285°Ct e 0 fro20 .m Witincreasine hth g respec t ively. energy capacitVVER-44e th f yo 0 load (320 eff.days a )necessit v The WER reactivity margin is compensated with a boric acid arises to compensate an excessive reactivity margin by increasing dissolve primara n i d y coolant the boric acid concentration. Thicausy alteratioe ma sth e n a f no e operatinth sign i n g temperatur ea highe rangd an re startup Fas poweI td parran f to temperature ) effect f reactivito s n i y temperatur becomy ma e e necessary r load.Fo f sKolo P Unia aNP t3 emergency and transient conditions are compensated with mechanical conservative startup temperatur f e26O>o s choseCwa whict na h «/at members of a control and protection system r The use of enhanced-enrichment (up to 4.4X) fuel in WER poses additional requirements to providing nuclear safety. TABLE I The fifth fuel load to the Kola NPP Unit 3 reactor included 78 Power and temperature coefficients of reactivity at energy assemblies enriche% Maximu4 4 o t dm reactivity margi 17.6Xs nwa o .T leve WER-44f lo 0 power load was experimental. therefore the concentration of boric acid in coolant under shutdown conditions was increased up to 16g/k«. In Exper iment Calculation this case subcriticality of a completely loaded reactor in cold 1. Transient fuel cycle condition with withdrawn mechanical e membercontroth d f an o sl Power, %/MW -1.00 10-» -0.92 10-» protection system was 8.3% (according to safety requirements subcriticality lest shoulno s e thadb n 2%) shoult .I d als notee ob d Temperature^, %/grad -1.85 10-* -1.77 that with an enhanced-enrichment fuel in a reactor, a differential efficiency of boric acid in its absolute value reduces by more than Transien. 2 t fuel cycle 25% in comparison with its value for a design regime of WER-440 Power, %/MW -1.44 10-» -1.27 1C-» refuelling This effect favours safety in the case of accidental Temperature, %/grad -2.42 10-* -2.63 10-* dissolvin boron-containinf o g g coolant with pure t i water t bu , aggravate e reactoth s r maneuvering becausreductioe th e f th o en i n 3. Transient fuel cycle rate of reactivity release by water-exchange when it is necessary to Power, %/MW -1.26 10-» -1.21 10-» compensate non-stationary xenon poisoning. Temperature %/grad -2.42 10-2 -2.85 10-2 Further calculations showed that the shutdown concentration of boric acid at the level of 13 27g/kg provides a sufficient 4 Stationary fuel cycle subcriticality Taking into account experimental and calculational Power,%/MW 0 -1 lO-3 1 19 IQ-3 error shutdowa s n concentratio recommendee b n t 14g/kca no ] r [2 gfo d Temperature , %/grad -2 80 10-2 2 88 10' fuel loadings having the cycle 350-380 eff days long ,1.0 0,5 230 240 250 TemperatureC , Fig. 3 Water and fuel coefficient of reactivity vs temperature with boron acid concentration of 8.7 gAg ——— calculation o experiment -14.0 - any position of a. control group over a control range is ensured negative [2]. The value f o sreactivit y coefficient r transienfo sd an t stationary (all fuel ig enriched to 4.4%) loads are shown in temperature TablTh - 1 e e dependenc f «fieo s show n Fig.3i n . 4.2. VVER-1000 e WER-100th In 0 cores e totath , l reactivity margis i n compensated by two systems: mechanical members of a control and T,*C protection system (61 members) in the form of bunches of thin absorbing components madboron-containinf o e a g n i materia C A 8 (1 l bunch borid )an c acid dissolve coolantn i d . With a 3-year fuel lifetime (fuel enrichment, is 4.4%) in VVER- 1000 reactors, to improve safety a part of reactivity margin is compensated by burnable absorber. Rods with burnable absorber are used in fresh assemblies during the first year of operation. In order to improve reactor safety, considered is an introduction of burnable absorber into the fuel (the use of uranium-gadolinium Flg.4 Core reactivity vs. average coolant fuel or pellets with deposited boron-containing material). temperature (Cb-0.3 gAg) memberl al ) controf a so protectiod lan n f burnablo e Thius se absorbe rt morge helpe o t snegativ e systems are inserted highest-worte th ) b h grou "stucks pi n "i reactivity coefficient with respect to coolant temperature at the upper position beginnin f fueo gl lifetime [4,5] alsd ,o increasan t o e numbeth e f o r CO 61.97.121 - number of members of control Ol contro d protectioan l n member orden i s o improvt r efficience th e f o y and protection system reactor emergency protection (e g with 97 members ot control and oxidatio d hydrationan smalls i n , oxide film thicknes s lesi s s 0> protection system a repeated criticality at the end of a burnup than 3 urn, cycl excludes i e wholde th eve ef i n highest-wort groud hro stucs pi h tuel has not undergone any essential structural change, column- in the upper position (Fig 4) like observedt grainno e sar ; - size changes of fuel elements are unimportant, 5 OPTIMIZATIO FUEF O N L ELEMEN ASSEMBLD TAN Y DESIGD AN N - internal gaa pressure at the end of exposure does not excess MATERIALS IN WER REACTORS WHEN TRANSFERRING TO ENHANCED the coolant pressure (gas release is insignificant); FUEL BURNUP studies hav t foun signy eno an d s indicativ f impossibiliteo f yo further operation 5 1 WER-440 e tueTh l elemen assembld an t y designs were improve provido t d e reliable operatio tuet no l element d assembliesan switchinn si g from 5,2. WER-1000 3- to 4-year fuel cycle In switchin 3-yeaa o gt r WER-IOOO fuel cycle with burnup higher Here we mention some of these improvements- than 40MW day/kg and enrichment of 4 4X, with the aim to improve the - pressur heliuf eo m fillin tuel-clae gth d spac fuen ei l elements operational reliability of fuel elements the diameter of fuel s increasewa MPa5 combinatio0 n I .o dt fro1 m0. thit no s factod ran pelleted central holes was increased from 1.4 to 2 4 mm. This helped a central hole ot 1.2-2.0 mm dia in the fuel pellet provided the to reduce the fuel temperature and hence the gaseous product reduction of fuel temperature and hence of gaseous fusion product release d alsincreaso an t ,o efrea e space pressura t .A heliuf eo m release [8), inside a clad of 2 0-2 5 MPa, a sufficient stabilitv of fuel clad is provided by the end of the burnout cycle C9]. to provide a free travel of fuel column and to reduce the formatio gritsf no fuee ,th l pellets with facets were introduced: Besides numbea , stepf ro s takeswa improvo nt fuee clad th elan d quality. In particular, the moisture and fluorine content in pellets to compensat radiation-inducee th e d growt f ho fue l elements, was reduced e productio,th f ngrito s reducewa s d (pellets with the space between the fuel element upper ends and the protective facet applied!e w s admissibld ,an e maximal dept claf ho d defects swa assemble th gri f o d y m hea m s increase5 dwa 2 o t p du reduced down to 50 mem etc [9] A number of technological improvements was also introduced with Analysi commerciaf o s l fuel elemen clad an dt parameters showed the aim to improve the quality of fuel pellets and fuel elements reasonable conformity with technical requirement good san d stability the tolerances on the moisture and fluorine content in fuel pellets, [10] fuel column length and mass, fuel clad diameter etc were made closer It ig envisaged to continue optimization and improvement ot fuel element and assembly design and performance in the nearest future Comprehensive post-reactor studie n o VVER-44s 0 assemblies with the aim ot providing high operational reliability at a more 4-yead an operate - r3 n fuei dl cycle conditions indicated good state extended burnup The following improvements are expected to be of tuel elements and other assembly components Below we give some accomplished results obtained wit assembln a h y operate NVNPn i d P Uni durin3 t - g5 year cycles and achieved the average burnup of 49 MW day/kg [8] fuel elemen withoum m t 0 elongatiot10 o changint p nu g fuel assembly overall size This will increase a free space in a eas tuel clads preserv e th plasticite y margin, mechanical collecto a fue r lo r column height, propertie t diffeno o d sr from thos fuef eo l element assemblief so s operated for three one-year cycles. - introductio spacerso t r Z t .no application ot integrated burnable neutron poisons (uranium- improvements will require additional validation and appropriate gadolinium fue boron-compounr o l d depositio pellen no t surfaces) technical specifications on Safety. technology, safety, and economy issues relative to uranium- gadolinium-tuelled VVER-lOuO reactors are also covered in this country by the IAEA BAF program of coordinated studies. Some . EFFEC7 EXTENDEF TO D VVER-FUEL BURNUE INITIATH N PO L results obtained within the framework of this program are presented STAGE OF A NUCLEAR FUEL CYCLE in [5]. Post-reactor studies on VVER-1000 fuel elements, exposed in the e Th applicatio t o nenhanced-enrichmen t fuel assemblier to s MR reactor under relatively severe condition burnua 0 5 o t f o p s(u VVER-440 fuelling pose additionae sth l requiremen nucleao t r safety MW day/kg) showed thachange th t fuel-elemenn i e t geometric sizs ei in handling fresh fuels. To validate nuclear safety it is necessary insignificant oxidefile th , m thicknes fuee cooland th lan n so t side o providt e such condition f sfreso h fuel storag t whica ee th h ia smal e 5 fue« lyVmth l d )claan d preserve a sconsiderabl e subcriticality of not-less than 0.05 at any accident could be plasticity margin [11]. ensured [12] Post-reactor studies on 2- and 3-year operated WER-1000 fuel e freshpitca Th f h4/6-enricheo 4 , d fuel assembly lattice which assemblies confirmed the validity of particular selected fuel ensures nuclear safety was calculated by the change of an effective element and assembly design and parameters. multiplication factor in infinite lattices placed in water of Post-reactors data agree reasonably well with preliminary fuel different density with different content of boric acid [13] element calculations performed by the RETR and PIN 0.4 codes. The racks for storing tresh fuel assemblies should be designed so that assemblie e sspaceb triangulaa n i d r lattic 5 pitca 22 t f ea ho mm, whic sufficiens i h ensuro t e safety. 6.VALIDATION (LICENSING EXTENDEN A F )O D VVER-FUEL BURNUP Similar means and conditions for storing fresh fuels are applied Early fuel load WER-100n i s -44d an 00 were experimentalld yan to VVER-1000 reactors. theoretically e validateordeo t burnuth t 29-3o r f p o u p2d MW.day/kg. Any further step to extending burnup required additional validation of WER-440 and -1000 switching to a 4- and 3-year fuel EFFEC8 EiCTENDEN A F TO R FUEUWE L BURNU FINAA N PO L lifetime, respectively, wit burnua h p averaged ovee fuellinron f go STAGE OF NUCLEAR FUEL CYCLE up to 40-42 and 40-45 MW.day/kg, respectively. Appropriate technical safety specifications were formulated and approved by State e effect Th extenden a f so d fuel burnu n transportatioo p d an n These specifications included e alsfollowinth o g basic fuel- long-term storag t speneo R nucleatWE r fuel e considerear s n i d related documents- design specifications, loop experimental results, detail within the tramework of the IAEA BCFAST" program ot calculation n fuel-elemeno s assembld an t y operational reliabilitv b y coordinated studies, whil issuee th e VVER-44f so 0 spent nuclear tuel th 4 ecodes 0. teste N , PI dneutronic d PETan R s calculations reprocessing (VVER-1000 spent fuel will be reprocessed by the RT-2 determinin conditioe th g f fueo nl operation (BÎPR-7. PERMAK etc.), plan w undeno t r constructio nI wer e delivered lase tth falo t l analysi operationan so l experience with fuecoresd an l . Meetin f IAEo gA Advisory Committe n Speneo t nuclear fuel handinn I g The possibility is investigated to extend the WER-fuel burnup to SO this section, the effects of fuel enrichment in U-235, assembly MW.day kg (5- and 4-year fuel cycles for VVER-440and 100O, capacity, burnu d operatiopan n condition n decao s v hea t o tstsen t 9 1 1.6 5.9S (max) 1 6 7 2 i.6 3.94 (av. ) 1 6 3 2.4 5.95 (max) 2 5 4 2.4 3.94 (av. 1 2 5 3.6 5.94 (max. ) 3 6 3.6 3.94 (av. ) 3 Fig.5 shows the dependences of VVER-440 spent fuel decav heat undeconditione th r long-terf o s m storage. Tabl give2 e s datn ao I02 operation conditions, initial fuel enrichmen burnupd tan . 9 Calculation decan o s y heat were performed using into account only Ö 7 the products of U-235 and PU-239 fission (8 - and j'- radiations were 6 included). The contribution of actinides to heat release is 5 neglected. 4 REFERENCES 1. Abagya n. A-Eal Statu t .e developmend san t issue f Sovieso t Nuclear Power", Atomnaya Energia, v.69. No.,2, 1990, 67-79. 2. Proaelko v. "SwitchinV.Nal t e . f gWER-44o o enhancedt 0 - ennchment fuels . Atomnaya Energia, v.69. No.2. 1990, 81-87 3 Proselko Voznesensk. N V R v WE Fue A V Increasle i th f o e I01 Burnu . pt ProTopicaIn c l R MeetinFueLW l n o gPerformance . 0 CO 0 40 100O 80 0 0^0 2000 3000 Storage time, days Avignon, France. 1991. vol.1. 252-255 4 Astakhov S A Proselkov VN et al Optimization ot WER BURNU SAFETA S PA Y PARAMETER RFO cores to improve tuel utilization and reactor safetv . VANT. ser HANDLING TRANSPORT PACKAGES Fizika yadernykh reaktorov . No 2. 1989, 71-75 STORAGD AN E FACILITIER SFO 5 Onutnev V D , Proselkov V N .Simonov V D Safetv assurance SPENT NUCLEAR FUEL and economy improvement in water-cooled Soviet reactors with uranium-gadoliniu F PK1meetint mBA 1s t fue .o e grepor, lth t a t V.l. KULIKOV, M.S. TIKHONOV, IAEA, Vecha. 1990 G. Ya. PHILIPPOV, A.I. IVANYUK Ail-Union Project and Research Institute 6 Pazdera F Unufnev V D , Proselkov V N et al Incentives for of Complex Power Technology, Extende R d WE ReactorBurnu r fo , spPro n co Fue lM IAETC A Leningrad, Union of Soviet Socialist Republics Performanc Higt a eh Burnu r Watefo p r Reactors 8 Jun5- .e 1990, Sweden, IWGFPT/36 Presented V.D.by Onufriev 7 Water Reactor Extended Burnup Study (WREBÜS). IAEA TRS. to be pub 1Ished Abstract 8 Dubrovin K P , Proselkov VN et al Improvement ot Four types of casks used in the Soviet Union for spent fuel commercial VVER-440 fuel element design with the aim to extend fuel transportation, namely TK-6, TK-10, TK-1 TK-1d considere1e an 3ar n I d burnup . Proceedings of I V Kurchatov Institute, v.l. 1990. 48-51. this paper in connection with issues arising from burnup Increase. It s beeha n shown that nuclear safety (undercriticallty d permissibl)an e 9. Bibilashvili Yu K Relation between fuel element performance radioactive content lost from a cask are not limiting factors in burnup increase sense. Residual heat dissipation and biological shielding of and desig d requirementan n s pose n commerciao d l fuel elementd san th e issue caskth e ssar relevan o burnut t p Increas d discussean ee th n i d component . st Pro o IAEcA Semina n o fuer l qualitv assurance. paper. Some attentio gives i n o ensurint n g safet f highlo y y burnt spent fuel discharged from WER and RBKK type reactors. Karlsruhe. Germany, June 19-21, 1990. 41-63 10 Shcheglov A S . Proselkov V N , Enin A A Statistics data on design and technological parameters of WER-1000 fuel elements . Preprint IAE, No 5334/4, 1991 1. Introduction 1 1 l a ApplicatioDubrovit e P K n experimentaf o n l WER-1000 fue l R M reacto elemente th Atomnay. rn i s . 5 ao N Energia . 62 v . Specific development of nuclear industry in most countries 1987, 312-317 of the world demands: - increas n initiai e l fuel enrichmen r operatinfo t d an g 12 Nuclear Safety Standards for NPP reactor plants, 1990 projected NPP; 13 l a AndrushechkNuclea t e rA S Safeto v Assurancn i e - increas fuen i e l burnup durin P operation-gNP , operating enhanced-ennchment tuel VVER-44n i s 0 reactor Atomnay. s a - increase in SNF storage capacity; Energia, v 69. No 2 1990 84-87 - justification of dry method for SNF storage. In overcoming the first two problems it is necessary to estimate whethe e numbeth r f existino r g cask capabls i s f eo transporting such fuel with enhanced characteristics from NPP t reprocessinoa casa g f sitkI doeeprovidt sno safete th e y level required,i s necessari t o considet y r transportation tCoO methodology beihg suitably changed. o Safe transportation of NPP spent fuel is generally provided o witfollowine th h g characteristic caska f so : Table 1 - efficient and reliable radiation shielding; Spent Fuel Package Characteristics - secured sub-cnticalit packagesf yo ; - abilit r fuefo yl assembly decay heat removal; I f TK-13 - permissible losses of volatile fission products. ! Cask identification TK-6 J TK-10 ! TK1 1 - General approach, to complete the cask loading with spent ! Fuel 1 1 ! fuel m case of fuel parameter values surpassing those assumed f Type of fuel PWR j PWR EiW ! LMFBRR ! PWR BWR ! (WWR-440)!(WWR- (RBMK-! (BN! - (WWR- (ACT- in cask designing, is as follows: j » 1000) 1000)! 600)» 1000) 500) j f t t Determin e th above whic ef o hfou r characteristics ! Initial enrichment, j i » represents a limiting factor. Then either select the most ! wt% U-235 3. 6 ! 4. 4 2 1 33 » 4.4 2 ! Number of assemblies 6 ! 30 ! 5 3 ! 51 12 12 suitabl a cask(reducin er loafo d e numbeth gf fueo r l t Average burnup, GWD/MTU 20/28 ! 42 20 ! 80 » 42 15 «.Cooling time, years 3 ! 3 313» 3 3 assemblies, increasing their cooling time) or change ! Maximum decay heat.kW 8/12 ! 13 10.3 ! 10.7 1 20 20 transportation pattern(replacing the cooler in a cask or using f Loaded fuel.tu/pack 3. 6 ! ! 2 . 0 6 1. ! 5.9 5.2 4.8 ! neutron-shielded casks for transportation). f Cavity t ? f f f } t ! Diameter, cm 0 10 . « 1475 . 148.5 ! 1=£ 2. Cask Description ! Length, cm 348 ! 503 379 » 49 5.5 ! Coolant in a cask gas/ water t wate! r s ga ga ! s gas gas f Pressure in a cask.atm 66/2.51 5 2. ! « C ! 2 1.66 1.66 Ther foue ear r type caskf so s used frequentl spenP NP tr yfo I | ! f ! Shielding: j ! t fuel transportatio e th Sovie n i t n Union, e thear y ] f j i TK-6, TK-10, TK-11 and TK-13. «.Gamma shield material stee . stee« l l steel t steel «.Thikness of shielding, i i The TK-6 cask is designed for transporting by rail thirty f cm 36 1 38 36 » 36 PWR assemblies from WWER-440-type(as wels a BK-50-typel ) ! Gamma dose rate 2m Î ! 1 'from side surface. i 1 f f reactor plants. It is a thick-walled forged flask weighing f mrem/h 2 f 3 4/ 4 ( 3 » 5.6 J 1.6 ! Neutron shield ' eth. l t ethylene more than 78 tons. The cask walls, bottom and lid protect f material ! - îglyco! l- 1 - glycol «Thickness of t ! ! f against ionizing radiation speciao n , l neutron shielding being 2 1 ! 1 - shielding ! , cm - i - » 12.7 provided. Fuel decay hea removes i t naturay db l convectiod nan f Neutron dose rate 2m f » | t 1 ! fiom side surface, ! ! ' i i radiation from the cask outer surface with many ribs. The4 . 0 ! ! 4/26 . ! mrem/h 3 ! 1 ! 1 I 0. 02 t l t t t 1 inne e rb fille case cavity th ma k d f o y withr o eithe s ga r » ! I Overall ! t ( water. ! characteristic1: ? 1 1 1 i ? « f ( t i e TK-1s intendei Th 0 o t transporR d PW y x raib tsi l ! Outside diamerter.cm ! 219.5 ! 200 219.5 ' 229.5 ! Length of cask, cm 1 410.5 ! 613 445. 5 ! 600.0 assemblies from WWER-1000(or WWER-50 d ACT-500)-typan 0 e t Mod f Î surfaco estee smootf le h steel fins ! smooth reactor. It is manufactured as a thick-wall forged steel ! ! fins (surface i surface ».Cask shape Î vertical Ihonz. vertical ' horizontal flask, on the outer side of which there is a liquid neutron Î (cylinder fcylmd. cylinder 1 cylinder ! ! Transport means Î r. car « r. car railway cai ! i ail way car ! shielding of ethylene glycol (67%) mixed with water. Residual 'Empty weight, t »76.5 ! 84 ! 86.5 " ! 86 6 . b10 « 106 decay nea removes i t naturay db l convectio d radiationan n from ! l the cask smooth outer surface. The TK-1 usee 1r raib casfo dy l kma transportatio fiftf no y Table 2 -one BWR assemblies from RBMK-1000-reactor or chirty-five Radiation Doses Outside Shieldin f Caskgo s with Fuel Loaded LMFBR assemblies from BN-type reactor alss i ot I usefu. r lfo with Maximum Design Burnup transporting fuel assemblies from WWER-440 and RBW.-1500-type reactors. This casdesigy simila s b TK-e i kd e th 6nan on o t r ï ) ! I f ! Cask identification ! TK-6 1TK-10 TK-1! TK-11 3 safety parameters. — — j f - — - • " I t — - - — — — 1f- _ - .- f! - - ! ! t 1 | The TK-13 cask is design for railway transporting twelve R PW Typ ! ! fuef eo l R (U02PW ! ) LITERI R R !PW BW BWR PWR assemblies from WWER-1000 reactor or twelve BWR assemblies ! l(WWR-440)!(WWR- (RFMK- (BM (WWR! - - (ACT- 1 ! f 1000) 1000)! 600)! lOOO) 500) from the ACT-500-reactor. The cask basic construction is a f t f ! Enrichment,wt7.U-235 ! 3. 6 t 4. 42.0 33.0 4.4 2.0 forged steel flask -with external neutron shieldint g of » ethylene ! glycoK67%)-and-water mixture. Decay hea dissipates i t d froma ! Cooling time, years ! 3 t 3 3 3 3 3 f ! ! smooth outer surface of the cask by natural convection and Coolan! transporr tfo gas/water! t ! water g-as gas gas gas ! ! ! radiation. »Maximum burnup, GWd/t 0 U? 20/45 ! 0 20 150 50 17 Some technical characteristic caske e th give ar sf n s o i n ! ! t lr! -quantu! m dose rat ! e Tabl. e1 ! ' at Em distance « 2/1 ! 0. 34 1 2.5 3.1 1.6 ! ! 1 ! Neutron dose rate ' I 2t ma distanc! ! 8/7. 0.05! e 5 4 0.3 4.3 0.02 3. Casks Design Potentialities 1 Heat output for cask,! ' ! kW ! 8/12 ! 13 10 3 10. 7 ! 20 ! 20 Accordine 'Statth eo g t Bodies' requirementr fo s supervision over safe transportation, package criticality questions should be decided upon without burnup being taken into consideration (i.e r fresfo . h fuel). Therefore, nuclear shielding, otherwis e numbeth f eo assembliere b o t s safety provision is not a limiting factor for the most of transporte reducede b y dma . containers designed for maximum fuel enrichment. Biological shielding of the casks being considered is The casks have a sufficiently reliable leak-proof system designe considerablo t d e allowance accountine b casa o r kt gfo eliminating volatile radioactive isotopes losses (both under loaded wit e maximuhth fuef o ml burnup whic possibls i h e normal transport condition emergencn i r o s y situation),that with reactor thif so s type. Tabl showe2 s ionizing radiation surpass permissible value casn i sf loadin eo e casth gk with dose awaratem y2 sfro m side surfac casa f keo fully loaded fuel promising the maximum burnup. Thus, permissible with fuel of maximum burnup, cooled for the design time-period radioactive content lost from a cask is not a limiting factor e temperaturTh f o fuee l assembly elements claddine th n i g either transport shoul . exceet C Thi no d0 s35 dput requiremente sth s With burnup rising abov average eth e valuese showth nm for the system of fuel assembly decay heat removal, and it is Tabl e1 neutro n sourc spenf o e t fuel drastically increases. just this condition thas oftei t a nlimitin g facton i r Furthermore, dose rate outside a cask is likely to exceed the determining the cask loading Thus, not all the loadings shown regulatory limits adopted thin I . s case, transportatioe b y nma in e feasiblee removaTablar th 2 ei l fo ,syste m does riot donwater-fillea n i e d cask providin r additionalfo g neutron provide value f o temperaturs e adopte regulatione th n i d r sfo o packages. For example, with the TK-13 cask fully loaded and ca transportee nb m dTK-6 water-fillee , th onl n i y d caskd ,an M WWER-1000-type fuel being cooled during up to 3 years, its its full loading is not permissible according to both nuclear exceeo t burnut 2 GWd/tUd4 no s n i pelevateA . d pressurf o e safety and radiation protection requirements. The problem can coolane casth kn i t cavity whic s expedieni h t witno t h be solve y replacinb d f fueo outee lth gw rassembliero s with standin e worsgth e condition a cas r k fo s leak-tightness their imitators which will serve as an additional shield and secured s accountei ,n effora r advanco t fo decaA dF n i ey supersed excess water case Th .k loading thin i , s case, reduces heat removal. fuet8 o1 l assemblies. In some instances, transportatio fuelf no , coole r lesdfo s than the design period,is required. Such conditions can arise 4. Opportunities Taken e froth Modifiem d Transport Methodology when the schedule for fuel dispatch to the reprocessing plant is changed or when the NPP is decommissioned ahead of schedule In accordance wit advancee hth d progra r safmfo e operation Armeniae th s a n NPP,for example. RBMK-reactof o r power plants, maximum fuel enrichmens wa t As for TK-6, the system for optimizing loading of cask increased from 2.0% to 2.4% with possible resultant growth in baskets is developed. The system is based on sorting out the fuel burnu abouo pt GWd/tU8 t2. . Transportatio sucf o nh fuel fuel assemblie e loadeb caska o t n si d , with burnup below 6wy/tU0 4 r o reducin, g their numbe r loadingfo r n I . in the TK-11 casks necessitates changes in the transport winter-time transportation, an increase in total release up to procedure: 1. For securing nuclear safety and pressure permissible in a cas r k pe bein W k 1insteaW 5k g2 1 full f do y loaded, proves the cask cavity under emergency conditions, transportation tpossiblee ob . Similar recommendations for transportation of WWER-type fuel shoul made db gas-fillen I e d casks only; with a relatively shorter cooling period as compared to a . Meanwhile2 , radiation shielding become e principath s l designed one, are developed to fit the TK-10 and TK-13 casks. limiting factor. Thus, fuel with initial enrichment of 20% and Accordin e experimentath o gt l data case th k, shiel wels i d l burnup of 226Vd/tU can be transported only after being cooled withi safete th n y e limitinmarginsth d an ,g facto a cass i rk for not less tnan 8 years. As to fuel enriched initially to design capacity for decay heat to be dissipated. For example, 2.4% and burnup of 28.5 GWd/tU,dose rate outside the shielding assemblies initial fuel enrichment below 4. 42, which are used exceeds standard values, even though the fuel is cooled for 15 under leactor start-up conditions without extended burnups, years thin I . s casnumbee eth fuef ro l assemblies loaded into transportee b n ca d when fuel cooling tims shortei e r than TK-11 casks is to be reduced;for example, it is possible, designed. Furthermore, radiation safety reauirements are without breakin safete gth y standards loao t 5 ,assemblie1 d s satisfied if decay heat from all the assemblies m a TK-10 being cooled for 8 years instead of the planned 51 assemblies cas TK-1a thad k an m doet excee3 t W k cask-2s no 3 1 d . 0kW loade caska morr n Fo i de. extended burnu w caspne fuek a l should be developed or a reactor storage built. To improve the economy, some nuclear power plants with VWER 5. Increasing Vaste Fuel Storage Capacity -440 reactors intend to use fuel with increased 4.4% enrichment. In this case, as seen from the experience of This proble s becomha m e more acut connection i e n wita h Kolskaya NPP, the fuel burnup can reach 50 GWd/tU. Such fuel decision take abandoo nt e reprocessinnth f go RBMK- fype spent fuel and, consequently, its long-term storage in quantities 5) carrying out base experiments for verifying a exceeding capacities availabl presentt A e , interim storage criticality analysis program and neutron-physical constants' densification is suggested to be doubled. The principal library available with regar o t fissiod n productd an s limiting factor for a design and densification rate to be actim des being present m burnt-up fuel. chosen is a nuclear safety requirement. The problem can be solved with using solid neutron absorbers, suc s basketa h s REFERENCES made of steels with boron, gadolinium etc. admixed This way is rather widely used, though involving considerable material 1. Kondratie .Kosare. N A v v U.A., Tikhono. S . N v and financial m recencosts , So t. timpossibilite th e f yo Storag Transportatiod ean Spenf o n t Nucleae rth Fuen i l solving the problem with burnup accounted for, is discussed. USSR. The Third International Conference on Nuclear Fuel Design studies conducted at the Physics-Energeties Reprocessin wastd an g e Management Registere B13(RS120): dN= . Institute, Obninsk, show that even though the 1986 decision on increasing initial fue X bein4 l. 2 genrichmen o t p u t . Makarchu2 .Morozo. F ,Tikhonok. T V . .BurlakovV . S . vN . A . V v considered, burnup taken as a nuclear safety parameter allows Justification of RBMK Reactor Spent Fuel Compacted Storage to overcom e probleth e densifief o m d RBMK-type fuel storage in Water Pools. International Seminar on Spent Fuel Storage without using solid neutron absorbers. Consideratio gives nwa n -Safety, Engineering and Environment Aspects, IAEA, Vienna, variouo t arrangemenA F s t pitches adopte storagn i d e projects Austria, 8-12,10,1990. and planne r theifo d r being desified: 250x16 230x11, 0mm , 0mm 250x8 , 125x110mm analysie , 115x11Th mm 0 . s0mm shows that 3. Basic Safet Physicad yan l Protection Rule r Nucleafo s r witfuey han l arrangement pitches m planneemergenc d an d y Material Shipment (OPBZ-83),Moscow,1984. situatios, a multiplication factor for assembly infinite lattice doeexceet sno value th d0.95f o e condition ,o n that Gernera. 4 P lSafet NP Rule r y fo sProvisio n fOPB-88), fuel is stored in canisters and its burnup goes over 10 GVM/tU PN AEG-1-011-89, Moscow, Atomenergoizdat, 1990. ( that correspond averago st ee first-yeaburnuth r pfo r fuel campaign). If the fuel is stored without canisters, the burnup should be not less than 21 GWd/tU. In its turn, using burnup as a safety parameter requires solvinr fo g thje following problems: nuclea) 1 r safety provisio storagr nfo fuef eo l assemblies with burhup being less than a designed one, Z) elimination of personnel errors when burnup being estimated, ) reliabl3 e evaluatio f burnt-uno p fuel isotopic content, 4) reliable estimation of burnup distributed along a fuel assembly, o CO IMPACT ON THE BACK END Reference policy of the Installed capacity OF THE NUCLEAR FUEL CYCLE whicf o h concerned countries. World USA France . BAIRIOH T FEX, Mol, Belgium Open cycle 52 % 31 '-. - Closed cycle - Abstract 45 \ 17 '', Extending burnup affects the characteristics of spent fuel to a larger Indefinite interim storage 17 \ extent tha simpla n e linear extrapolation. Increased alpha (and consequent decay ' heat generation d neutro)an n activitie predominane th e sar t features impactin' n go ' spent fuel management opee ,d close th botnan n i hd cycle options. Reprocesses ha U d FIG Bac. options1 .d ken expressed ,an base] 1 [ percentagen s do a worldwidf so e installed properties only marginally affected by the higher burnup. To the contrary, the proper- nuclear capacity f Decembeo s a , r 1991 (th100o t ep % u percentage d sincad et no o sd ties of Pu are strongly modified and higher Pu contents are required. However, the several countries have selected indefinite interim storage for part of their spent fuel to pro- vid delaea selectinr yfo mose gth t appropriate bacoption)d ken . time scale for the MOX fuel industry to cops with this challenging futura Is stretched ove ra lon g enoug manufacturinX MO he perioth r dfo g plant adaptee time b du o e t sn i d adequate th r anfo de data implementede basb o et . Span. 2 t fuel act1vitlea. For such 50 * increase In burnup, the radioactivity of fission products (fig.3) in the spent fuel (beta and associated gamma activity) Is Increased by 36 Ï per kg U(fig.4) 1. Introduction. and nance decreased by 9 Î per kWh generated. To the contrary, tha radioactivity of acti- The impact of extending fuel burnup should be considered in the perspective nidas (fig.4 spene th tn ) i fual (alph d associateaan d gamm neutrod an a n activitied san botf opeo e th hn (i.e. final disposa f spenlo t fue s finaa l l e closewasteth d d)an heat generation) is increased by 250 '-. per kg U and hence by 140 * per KWh generated. (i.e. reprocessing and recycling of the fissile material) cycle, since each alternative Proper management of the actlnides becomes therefore an Important Issue. For Instance, it is equally used (worldwide reference th s )a e bacoptiod en k n (fig.1). ) ,1 base( n o d is unclear whether and to what extent the high alpha and neutron activity would deteriorate riske th ss A associated with spent fuel became predominantly influence actlnidese th y b d , the cladding during very long term storage of high burnup spent fuel. rather than by the fission products, after 10 to 30 years storage time (fig.2), it is fues a lu P Reprocessin takee th e slarges d th car an f d recyclineo U t an ge th f o g important to assess how much extended burnup affects radioactivity to be managed in the part of the actlnides stream. The Increasing quantities of Cm and Am in the spent fuel fuel cycle. This paper takes, as an example, a 50 \ Increase of discharge burnup of PWR Justif developmente th y s being contemplate o Implement d t specific management schemer fo s fuel, from 32r33 GWd/tU to 18-50 GWd/tU. those products (isolatio d transmutatioan n r fission)o n . flSSIO« PBOOUCI BETS( GATT»I U ACTIVIT Sr(-Yl80• ) YCi(-6*. 19 c .T l 137 15«u .E 5 123.I 13 l .C RISK FACTOR (1 = ore needed HEAVY neTH 237o »ETINIOeN S( . 1362««2*1U . » » C » .Î43MS .m • 235 U » u S) .P , produco t fuele eth ) U 233. Pu 241. Pu 239. Pu 240. Pu 242 8904 Pu 241 IV) 241 Np 237 4 10 1200 i/t 24 i/t »30 f/t INDEFINITE STO3RAGE SE>:BNT FIG. 3. The most Important Isotopes with long half-lives. g/tHM Cl/tHM DFOTOn 33 GWd/t 50 6Wd/t EWd/3 3 t GWd/0 5 t Sr-0 9 Y * 2 8a 480 590 140 000 170 000 99 Tc 210 Ka 900 1200 15 21 135 Cs 2.9 Ma 1100 1700 1 1.5 Cs-B7 13 a + a 0 3 1100 1700 100 000 150 000 154 Eu 16 a 4.3 8 5 800 11 000 10-2 FPs 10 a 3600 5200 250 000 340 000 I ————— 237 Np 2.8 Ma 0.3 0.4 10 l 000 100 000 10 000 000 YRS Pu («Û 7-38a 0k 9000 11000 3 800 8 000 241 Am 450 a 800 3 000 244 Cm 18 a 1 900 12 000 STORA.GE TIME ACTINIOES 10 a 6 500 23 000 predominan FIGo . Evolutio2 tw . e th f no t component riske th .f so O cn FIG. 4. Inventory of long life fission products and actinides. 3. Spent fuel storage and final disposal. a>O Operation Attribut« ( « ) C 2 ) For a 36 ^ increased discharge bumup, the residual heat of spent fuel is increased Transport mmfctr of FA« - 31 t - 30\ by 26 \ after a cooling time of 1 yr and by 49 * after 5 yra ( 2 ). Since the decay t BOL «nrlcnmcflt » 8 2 • * 9 2 o t 8 • heat is one limiting feature for spent fuel transportation, the storage period In the »torse» n/tac aftar 3 yr» «120 t t O H • spent fusl pools at the reactor site might have to be Increased considerably. For Instance, Shearing I nuflfcar of FAs • 31 \ - 30\ taking the decay heat generation versus cooling time per assembly from the presentation dissolution Ru-R «ft»t 6 yr 10 h3 r - 16 * - 50 t increasn a r fo dischargf eo l n deca e , th ) y. 3 e % perio burnu( 7 6 f do p mus extendee tb d admissible th f i r fues epe yr l hea W 4 assembly 1 Q t o BD t loa s i da r fro fro,o yr n5 m Solvent Pu - 18 * - 18 t Cn •110 l becomest i f i s fue100r yr pe l 08 W assemblyo t s yr .4 Consequently thin ,i s example, «xtrectlon • 140 t activity « 0 < • • 19 t the at-reactor-site storage capacity expressed in number of years of reactor operation could be reduced by respectively 60 t or 17 *., It might bscome limiting in power plants Pu convurslon heat (anaratlon - 6 * » 3 • moderato havint w lo g e spent fuel capacity. HWL vitrification fla«f - 10 t t 5 - The large increase of residual heat per kg U (250 * in the example of fig.4) . QuantitativFIG5 . e variation r kW-,pe h generated attributee th f ,o s Influencing reprocess- becomes also a limiting feature for direct disposal of spent fuel at waste sites where ing, when discharge burnu spene th f pto fuel increase GW-d/7 s4 o frot . m3 U t 3 both package weights and haat generation are Key issues (e.g. clay formations). It could ultimatel necessita y th lea o t d f d separatinactinideyo an u P e se th g froth bula f mth o k Burn-ups 34000 MWO/tU £5000 MWD/IU Cooling UNIT. U-23S:3.25V.) (HOT. U-235: 4.2%) weight of spent fuel. before 5 years 5 years reprocessing Composition Activity Composition Activity (Bq/gUl (Bq/gU) 4. Spent fuel storage and reprocessing. U-232 I«) 0.82 ppb eeo b 2.4pp 9 1989 takind Basepasa moran a g n t) ga do n e o publicatio2 recend ( e an ) ton 4 ( n U-234 (o() 0.0128 '/. 29180 0.019! % <3S54 a unit, the spent fusl corresponding to a same quantity of electricity generated (fig,5), U-235 (cfj 0.75V. 590 0.928 V. 734 extended burnup affects negatively spent fuel transpor e solventh d tan t extraction step U-236 let) 0/7 V. 10670 0.583 % 1056) at the reprocessing plant. All the other operations are drawing benefits from burnup ex- U-237 Iß,/) 5.5 10". 'V 165800 4.2 lO'3 •/. 126826 tensio a.oi n e hig.th h level vitrified waste quantity decrease h generatedKW r spe , U-23) , 8[d 98.77 V. 12250 98.47% 12213 53350 69051 Total 100V. 100V. 5. Reprocessed uranium. P 165800 126826 U-232/U-235 0.109 ppm m 0.2pp 7 U originatin p Re g from extended burnup fuel 0 ''(flg.B2 . a highe s )ha r reactlvlt) worth U fro thap mRe n standard burnup fuel (i.e t affected) valus no a fue It .s s I la e . FIG. 6. Reprocessed uranium composition and activity. GWd/tU 33 43 45 53 PU SgB EOL yrs6 yrs2 . , 0 EOL 0 2 yrs. 6 yr». EOL 6 6 6 Pu 236 12X10"6 nd 3,4X10~6 nd 15X10"6 nd 4.6X10" 1.9X10" 20X10" Pu 236 1.3 1.8 1.6 1.7 2.0 2.6 2.8 2.7 2.7 Pu 233 56.5 58.2 56. 2 56.2 52.6 55.3 55.3 55.3 50.1 Pu 240 23.2 22.7 22,7 22.7 24.1 23.1 23.1 23.1 21.2 Pu 241 13.9 12.2 11.1 3.2 14.7 12.6 11.5 9.5 15.2 Pu 242 4,7 5.0 5.0 5.0 6.2 6,3 6.3 6.3 7 Am 241 0.35 1.1 3.0 0.44 1 3. 1.1 0.51 Reference ( 5 ) ( 7 ) ( 7 ) t 7 ) ) 5 ( t 7 ) t 7 ) ( 6 ) ) 5 ( FIG. 9 Pu isotopic composition (percentage by weight) as a function of burnup of the original spent fuel (GW-d/t U) and of Pu age (storage time after reprocessing). EOL is the final discharge from the reactor core, i.e. some years before reprocessing. t~l «•) CM <"* OS C 4 <*- W) CN r « + n CD n o CM (71 O o c o e o ce UJ L3 107 o> ACTIVITY FACTOR U FUEL MOX FUEL Rep U FUEL ALPHA 1.8 1980' S sU % 2 3. COMMERCIAL SMALL SCALE DEMO HEAT GENERATION 1.Ü GWd/tH2 3 M Pu ex 30 GWd/tU GAMMA -<200 KeV 1.5 5 Î Put 0.2 - 1 MeV 1.1 GWd/tH3 3 M >• 1 MeV 1.6 1990's 4.1 * US 0 GWd/t3 x Pe uU LARGE SCALE DEMO NEUTRON 45 GWd/tHflt Pu % 7 U ex 30 GWd/tU 45 GWd/tHfl S U * 5 FIG. Increas10 .radioactivit u P f eo y when spenU t 45 GWd/tHM fuel burnu increaseps i GW-d/8 d4 froo t . U 3 mt 3 2000 's 5 '-S U . 5 GWd/t4 x e U u P COMMERCIAL GWd/tH0 6 M t 8 '-Pu . GWd/t0 3 x e UU 60 GWd/tHM 5,5 * US GWd/tH0 8 M 2010's SAME Pu ex 60 GWd/tU GWd/t5 4 x e U U Year Plutonium Equivalent Average In-reactor + 30 8 45 GWd/tHM 6 % US content Ü-235 content reload burnup life 0 t '-1 Pu . (approx.) (approx.) (MWd/tHM) (years) 2020 ' s SAME Pu ex 60 GWd/tU U ex 60 GWd/tU + 45 & 60 GWd/tHM 7 % US 1990 5 3.25 % 36000 t Pu 1% 2 % 5 1997. 5 3.7 % 42000 FIG. Extende12 . d burnup implementation time scales. > 2000 10 % 4.5-5 % 55-60000 5 FIG. Expecte11 . d improvement fueX l performancMO n si e [9]. base datn o d a presente basie th morf so n daO earliereacrecen. e ) th 5 - . t( r) dat1 ( a less negativ evea ean n ultimately positiv situatioa . a n which require extension sa f no tivity wort decreases I h e radioactivit0 '-..2 Th y b d y Increase (fig.6e th o mainls )t I e du y the experimental data baseOo). hard gammas emitte daughtey b d r product 232U f ,so which form non-volatile fluoridess .A The actual perspective to be faced is however much more progressive than what a result, the inconvenience Is essentially limited to the enrichment plant and to the hexa- might be deduced from equilibrium cycle values (e.g. fig.11] : Pu to bs incorporated In fluorlde transport containers whicn i * h those non-volatile decay product beine sar g depo- targey X an fue nO f tlo discharge burnup will manr ,fo y decades u fropreviouP e e th m,b s sited. They must be decontaminated more frequently thereby generating additional process generation of discharge burnups (fig.12). This will provide timely opportunity for the waste. industry to collect an adequate data base and to implement appropriate technical eolutlons Plutonium. 6 . to this challenging evolution of feed Pu and MOX fuel to be fabricated, on the grounds of lessons progressively learned from commercial operatio fabricatioe th f powef o no d ran n The net quantity of Pu produced decreases, as part of the generated Pu is plants with MOX fuels. burne sitn I d u (fig.7) radioactivite .Th (flg.8u P f yo ) increases however sharply (fig.9), Finally, it is not unlikely that the nuclear industry will have to utilize, in highee th duo rt e proportio f alphno neutrod an a n emitting Isotopes (fig.10 harf o d )an MOX fuel. Pu arising from the dismantling of nuclear warheads as the most efficient way gamma activity of the daughter products of Pu 236. The challenge to the MOX fabrication to demilitarize that Pu. Mixing It to Pu arising from high burnup spent fuel will result plant will not only result from this Increased specific radioactivity, but also from tho in completely annihilating the Increased radioactivity of the latter. higher Pu contents required for extended burnup MOX fuel (fig.11). These highe contentu resultine rP th d san g neutron spectrum hardenlg decrease 7. Final remark. the capture effecth f o te résonance 238U resultt f voie .sI o th d n coefficienI s t becoming Altogether nucleae ,th rinstitute0 IndustrS R d takine an y sar necessar e th g y steps or preparing the appropriate initiatives to cope in due time with the back-end DIFFERENCES APs FPs ACTINIOES consequences of extending discharge burnups. however, is t I , observed (fig.13) that large differences exis predictinn i t e th g >-30 •-, Zr 93 Mo 96 Pu 239 quantity of each isotope in high burnup spent fuel. This might mislead the développera in preparing for the future. Opportunity should be taken of available high burnup U and MOX Xe 128 fuel buildo t s oncet ,a n experimentaa , l data base whicn ,o h predictive e codeb n sca Sm 147 benchmarke future th r e fo dburnu p range interestf so . Eu 156 REFERENCES ^>20 '-, Cs 135 Cm 244 ( 1 ) - Nuclear Powur. Nuclear Fuel Cycle and Waste Management : Status and Trends 1991. Part of the IAEA Yearbook 1991. o Predictiv. 13 FIG. e capabilit activatiof yo n products, fission product d actinidesan s content STI / PUB / 092. CO of fuel at high burnup (examples). ISBN 92-0-179291-3. Compatibilit- ) Z ( Extendef o y d Burnup with French Fuel Cycle Installations. . OemoullnP Saverot- P $ s . IAEA ftGCI " Impact of Extended Burnup on the Nuclear Fuel Cycle ", Wien Dec5 . 2- .. 91 Effect- ) 3 Extendef so ( d Nucleae Burnuth n o pr Fuel Cycle. P. Lang Ibidem. ( 4 ) - ANS topical meeting: "LWR Extended Burnup Fuel Performance and Utilization". Williamsburg, April 82. ( 5 ) - Recycling. R. Cayron 4 E. Crispino Genève, EN86 C . Jun. 86 e ( B ) - Pu Fuel - An Assessment, OEC NEAD/ . 1989. ISBN 92-64-13265-1. (71- (1ELOX: Impac Materiaf o t Operationad lan l Constraint Desige Larga th f n no o se Capacity Plant. Y. Couty S H. Lorenzelli. Presentation at the IAEA TCP! on "Recycling of Pu and U in WR Fuels", Cadarache, 13-16 Nov. 89. (81- MOX Fuel Fabrication Technology at BELGQNUCLEAIRE and CFCa and further Developments for MELOX. J.M. Leblanc Balriot. H , Lorenzell. ,R J.L8 i . Nigon Special Issu Nucleaf o e r Technology (1991). ( 9 ) - Recycling Developments in France. H. Simon. U Inst. Annual Symp. London Sept7 . ,5- .90 ( 10 ) - Analysis of PROTEUS Phase II Experiments in Support of Light Water High Conversion Reactor Design. R. Chawla et al. Kerntechnik 57 (1992) N°1. LIS PARTICIPANTF TO S FINLAND S Kelppe Technical Research Centr f Finlaneo d ARGENTINA Nuclear Engineering Laboratory 8 20 x Bo PO CasariA J o CNEA, Centro Atomico Banloche 840 Carlon 0 Sa Bariloch e sd e SF-02150 Espoo BELGIUM FRANCE H Bamot FEX P Desmoulins Electricit France éd e Lijsterdreef 24 DPT/Service des combustibles B 240l 0Mo 23 bis, avenu Messine ed e F-75384 Pans Cedex 08 C Delbrassine CEN/SCK Boeretang 200 QuaiC J s NUSYS B-240l 0Mo 14, rue du Printemps F-75017Pans H Duflou BetgatoA 5 V mN avenue Ariane7 GERMANY B-1200 Brussels M Peehs Siemens KWU DruennH e Tractebel Postfach 3220 boulevard du Régent D-W 8520 Erlangen B-1000 Brussels INDIA VandenberC g Belgonucléaire Chamu d e Mare ru pd 5 s2 H D Purandare Bhabha Atomic Research Centre B 1050 Brussels Theoretical Physics Division Trombay, Bombay 400085 CANADA JAPAN J Griffiths AECL Research Chalk River Laboratories A Toba Tokyo Electric Power Company Chalk River, Ontario KOJ 1JO 1-3, Uchisaiwao ch i 1 chôme, Chiyoda ku CHINA RUSSIAN FEDERATION Z Liu Shanghai Nuclear Engineering Research and Design Institute OnufrieD V v Scientific and Research Institute 8 8 3 23 x Bo PO for Inorganic Materials 200233 Shanghai St Rogov5 123060 Moscow CZECH REPUBLIC VProselkoN v KurchatoV I v Institut f Atomieo c Energy F Pazdera Nuclear Research Institute PI Acad Kurchtov, 1 250 68 Rez nr Prague 123182 Moscow ^ UNITED KINGDOM ro K. Hesketh BNFL, Springfield Works SalwickR , PrestoOQ 4 nPR D. Glazbrook Nuclear Installations Inspectorate St. Peter's House Stanley Precinct Bootle, Merseyside C.H. Walker Nuclear Installations Inspectorate St. Peter's House Stanley Precinct Bootle, Merseyside UNITED STATES OF AMERICA . LanP g Offic f Reactoeo r Deployment United States Department of Energy Washington, DC 20545 INTERNATIONAL ATOMIC ENERGY AGENCY i O . N Divisio f Nucleano r Fuel Cycle and Waste Management International Atomic Energy Agency Wagramerstrasse 5 0 10 P.Ox Bo . A-1400 Vienna, Austria . SukhanoG v Divisio Nucleaf no r Fuel Cycle (Scientific Secretary) and Waste Management International Atomic Energy Agency Wagramerstrasse5 0 10 P.Ox Bo . A-1400 Vienna, Austria 93-00492 base datn o d a presente basie th morf so n deO earlierecen. reace ) th 5 - t, ( r ) dat4 ( a less negativ eved ean n ultimately positiv situatioa , e n which require extension sa f o n tivity worth is decreased by 20 4. The radioactivity Increase (fig.6) Is mainly due to the the experimental data base{f0). hard gammas emitted by daughter products of U 232, which form non-volatile fluorides. As The actual perspective to be faced is however much more progressive than what a result, the inconvenience is essentially limited to the enrichment plant and to the hexo- migh deducee tb d from equilibrium cycle values (e.g e incorporate.b fig.11o t u P ]; n i d fluorlde transport containers, in which those non-volatila decay products are being depo- MOX fuel of any target discharge burnup will, for many decades, be Pu from the previous sited. They mus decontaminatee b t d more frequently thereby generating additional process generation of discharge burnups (fig.12). This will provide timely opportunity for the waste. industr colleco yt n adequatta e dat implemenao t bas d ean t appropriate technical solutions Plutonium. 6 . to this challenging evolution of feed Pu and MOX fuel to ba fabricated, on the grounds of lessons progressively learned from commercial operation of the fabrication and of power The net quantity of Pu produced decreases, as part of the generated Pu is plants with TOX fuels. burned in situ (fig.7). The radioactivity of Pu (flg.8) increases however sharply (fig.9), t unlikelFinallyno s i t y,i tha nucleae tth r industry will hav o utilizeet n ,i highee th duo rt e proportio f alphno neutrod an a n emitting Isotopes (fig.10 harf o d )an rtOX fuel, Pu arising from the dismantling of nuclear warheads as the most efficient way gamma activity of the daughter products of Pu 236. The challenge to the MOX fabrication to demilitarize that Pu. Mixing it to Pu arising from high burnup spent fuel will result plant wilonlt no ly result from this Increased specific radioactivity alst ,bu o froe mth in completely annihilating the increased radioactivity of the latter. highe contentu rP s requlro r extendefo d d burnu fueX pMO l (fig.11). These higher Pu contents and the resulting neutron spectrum hardenlg decrease 7. Final remark. the effect of the capture résonances of U 238. It results in the void coefficient becoming Altogether nucleae ,th rinstitute0 IndustrS R d takine an y sar necessare th g y step preparinr so appropriate gth e initiative e tim du copo e st n witI eback-ene th h d DIFFERENCES APs FPs ACTINIOES consequence f extendinso g discharge burnups» , howeverIis t , observed (fig.13) that large differences exis predictinn i t e th g >30 •-. Zr 93 Mo 96 Pu 239 quantity of each isotope in high burnup spent fuel. This might mislead the developpers in preparin futuree th r .fo g Opportunity shoul takee b d f availablo n e higX h MO burnu d an U p Xe 128 fuel buildo t s oncet ,a n experimentaa , l data base whicn ,o h predictive e codeb n sca Sm 147 benchmarked for the future burnup ranges of Interest. Eu 156 REFERENCES ^> 20 '-, Cs 135 Cm 244 (11- Nuclear Powar. Nuc)ear Fuel Cycle and Waste Management : Status and Trends 1991. Pdrt of tlie IAEA Yearbook 1991. o Predictiv. FIG13 . e capabilit activatiof yo n products, fission product d actinidesan s content 092/ B .PU ST / I CO of fuel at high burnup (examples). ISBN 92-0-179291-3. ( 2 ) - Compatibility of Extended Bumup with French Fuel Cycle Installations. Oemoulln. P . SaverotP S s . Impac" IAEM f ExtendeAC Ao t d Burnue Nucleath n o p r Fuel Cycl, f e Wien, 2-5 Dec. 91. Effect- ) 3 Extendef so ( d Nucleae Burnuth n o pr Fuel Cycle. P. Lang Ibidem. (41 S topicaAN - l meeting:"LWR Extended Burnup Fuel Performanc Utilization"d ean . Williamsburg, Apri. 82 l ( 5 ) - Recycling. R. Cayron 4 E. Crispino ENC 86, Genève, June 86. ( 6 ) - Pu Fuel - An Assessment, OECD / NEA. 1989. ISBN 92-64-13265-1. MELOX- Î 7 : ( Impac Materiaf o t Operationad lan l Constraint Desige Larga th f n no o se Capacity Plant. Y. Couty $ R- Lorenzelli. Fuels"R W n Presentatioi .U n "Recyclind o e IAEan M th TC u AP t a nf o g Cadarache, 13-16 Nov. 89. (81- rax Fuel Fabrication Technology at BELGQNUCLEAIRE and CFCa and further Developments MELOXr fo . J.M. Leblanc Balriot. H , Lorenzell. ,R J.L8 i . Nigon Special Issue of Nuclear Technology (1991). Recyclin- ) 9 ( g Development Francen i s . M. Simon. U Inst, Annual Symp. London. 5-7 Sept. 90. Analysi- ) 0 1 PROTEUf o s ( S Phas ExperimentI eI Supporn i s Lighf o t t Water High Conversion Reactor Design. R. Chawla et al. Kerntechni 7 (19925 k ) N°1. LIST OF PARTICIPANTS FINLAND S Kelppe Technical Research Centre of Finland ARGENTINA Nuclear Engineering Laboratory CasariA J o CNEA, Centra Atomico Bariloche P O Box 208 8400 San Carlos de Banloche SF02150 Espoo BELGIUM FRANCE H Bainot FEX P Desmoulins Electricit France éd e Lijsterdreef 24 DPT/Service des combustibles B 240l 0Mo 23 bis, avenue de Messine F 75384 Pans Cede8 x0 C Delbrassine CEN/SCK Boeretang 200 J C Guais NUSYS B 2400 Mol Printempu d e ru , s14 F 75017 Pans H Duflou S A Beigatom N V avenue Ariane7 GERMANY B 1200 Brussels M Peehs Siemens KWU DruennH e Tractebel Postfach 3220 boulevard du Régent D W 8520 Erlangen B 1000 Brussels INDIA C Vandenberg Belgonucléaire Chamu d e Mare ru pd 5 s2 H D Purandare Bhabha Atomic Research Centre B 1050 Brussels Theoretical Physics Division Trombay, Bombay 400085 CANADA JAPAN J Griffiths AECL Research Chalk River Laboratories A Toba Tokyo Electric Power Company Chalk River,O Ontan1J J oKO 1 3, Uchisaiwai cho 1 chôme, Chiyoda ku CHINA RUSSIAN FEDERATION Z Liu Shanghai Nuclear Engineering Research Desigd an n Institute OnufneD V v Scientific and Research Institute 8 8 3 23 x Bo PO for Inorganic Materials 200233 Shanghai St RogovS 123060 Moscow CZECH REPUBLIC V N Proselkov I V Kurchatov Institute of Atomic Energy F Pazdera Nuclear Research Institute PI Acad Kurchtov 1 r Pragun z Re e 8 6 0 25 123182 Moscow ^ UNITED KINGDOM ro K. Hesketh BNFL, Springfield Works SalwickR , PrestoOQ 4 nPR D. Glazbrook Nuclear Installations Inspectorate . Peter'St s House Stanley Precinct Bootle, Merseyside C.H. Walker Nuclear Installations Inspectorate . Peter'St s House Stanley Precinct Bootle, Merseyside UNITED STATES OF AMERICA P. Lang Office of Reactor Deployment United States Department of Energy Washington 2054C D , 5 INTERNATIONAL ATOMIC ENERGY AGENCY i O . N Divisio f Nucleano r Fuel Cycle and Waste Management International Atomic Energy Agency Wagramerstrasse 5 0 10 x P.OBo . A-1400 Vienna, Austria . SukhanoG v Divisio Nucleaf no r Fuel Cycle (Scientific Secretary) and Waste Management International Atomic Energy Agency Wagramerstrasse5 0 10 x P.OBo . A-1400 Vienna, Austria 93-00492